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.

Photon transport in thin disordered slabs

Indian Academy of Sciences (India)

We examine using Monte Carlo simulations, photon transport in optically `thin' slabs whose thickness is only a few times the transport mean free path *, with particles of different scattering anisotropies. The confined geometry causes an auto-selection of only photons with looping paths to remain within the slab.

Continuous Energy Photon Transport Implementation in MCATK

Energy Technology Data Exchange (ETDEWEB)

Adams, Terry R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Trahan, Travis John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sweezy, Jeremy Ed [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nolen, Steven Douglas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hughes, Henry Grady [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Pritchett-Sheats, Lori A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Werner, Christopher John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-10-31

The Monte Carlo Application ToolKit (MCATK) code development team has implemented Monte Carlo photon transport into the MCATK software suite. The current particle transport capabilities in MCATK, which process the tracking and collision physics, have been extended to enable tracking of photons using the same continuous energy approximation. We describe the four photoatomic processes implemented, which are coherent scattering, incoherent scattering, pair-production, and photoelectric absorption. The accompanying background, implementation, and verification of these processes will be presented.

Penelope - a code system for Monte Carlo simulation of electron and photon transport

International Nuclear Information System (INIS)

2003-01-01

Radiation is used in many applications of modern technology. Its proper handling requires competent knowledge of the basic physical laws governing its interaction with matter. To ensure its safe use, appropriate tools for predicting radiation fields and doses, as well as pertinent regulations, are required. One area of radiation physics that has received much attention concerns electron-photon transport in matter. PENELOPE is a modern, general-purpose Monte Carlo tool for simulating the transport of electrons and photons, which is applicable for arbitrary materials and in a wide energy range. PENELOPE provides quantitative guidance for many practical situations and techniques, including electron and X-ray spectroscopies, electron microscopy and microanalysis, biophysics, dosimetry, medical diagnostics and radiotherapy, as well as radiation damage and shielding. These proceedings contain the extensively revised teaching notes of the second workshop/training course on PENELOPE held in 2003, along with a detailed description of the improved physic models, numerical algorithms and structure of the code system. (author)

Modified Monte Carlo procedure for particle transport problems

International Nuclear Information System (INIS)

Matthes, W.

1978-01-01

The simulation of photon transport in the atmosphere with the Monte Carlo method forms part of the EURASEP-programme. The specifications for the problems posed for a solution were such, that the direct application of the analogue Monte Carlo method was not feasible. For this reason the standard Monte Carlo procedure was modified in the sense that additional properly weighted branchings at each collision and transport process in a photon history were introduced. This modified Monte Carlo procedure leads to a clear and logical separation of the essential parts of a problem and offers a large flexibility for variance reducing techniques. More complex problems, as foreseen in the EURASEP-programme (e.g. clouds in the atmosphere, rough ocean-surface and chlorophyl-distribution in the ocean) can be handled by recoding some subroutines. This collision- and transport-splitting procedure can of course be performed differently in different space- and energy regions. It is applied here only for a homogeneous problem

Development of a Monte Carlo software to photon transportation in voxel structures using graphic processing units; Desenvolvimento de um software de Monte Carlo para transporte de fotons em estruturas de voxels usando unidades de processamento grafico

Energy Technology Data Exchange (ETDEWEB)

Bellezzo, Murillo

2014-09-01

As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo Method (MCM) has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this thesis, the CUBMC code is presented, a GPU-based MC photon transport algorithm for dose calculation under the Compute Unified Device Architecture (CUDA) platform. The simulation of physical events is based on the algorithm used in PENELOPE, and the cross section table used is the one generated by the MATERIAL routine, also present in PENELOPE code. Photons are transported in voxel-based geometries with different compositions. There are two distinct approaches used for transport simulation. The rst of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon ignores the existence of borders and travels in homogeneous fictitious media. The CUBMC code aims to be an alternative of Monte Carlo simulator code that, by using the capability of parallel processing of graphics processing units (GPU), provide high performance simulations in low cost compact machines, and thus can be applied in clinical cases and incorporated in treatment planning systems for radiotherapy. (author)

ITS, TIGER System of Coupled Electron Photon Transport by Monte-Carlo

International Nuclear Information System (INIS)

Halbleib, J.A.; Mehlhorn, T.A.; Young, M.F.

1996-01-01

1 - Description of program or function: ITS permits a state-of-the-art Monte Carlo solution of linear time-integrated coupled electron/ photon radiation transport problems with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. 2 - Method of solution: Through a machine-portable utility that emulates the basic features of the CDC UPDATE processor, the user selects one of eight codes for running on a machine of one of four (at least) major vendors. With the ITS-3.0 release the PSR-0245/UPEML package is included to perform these functions. The ease with which this utility is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is maximized by employing the best available cross sections and sampling distributions, and the most complete physical model for describing the production and transport of the electron/ photon cascade from 1.0 GeV down to 1.0 keV. Flexibility of construction permits the codes to be tailored to specific applications and the capabilities of the codes to be extended to more complex applications through update procedures. 3 - Restrictions on the complexity of the problem: - Restrictions and/or limitations for ITS depend upon the local operating system

Modelling of an industrial environment, part 1.: Monte Carlo simulations of photon transport

International Nuclear Information System (INIS)

Kis, Z.; Eged, K.; Meckbach, R.; Voigt, G.

2002-01-01

After a nuclear accident releasing radioactive material into the environment the external exposures may contribute significantly to the radiation exposure of the population (UNSCEAR 1988, 2000). For urban populations the external gamma exposure from radionuclides deposited on the surfaces of the urban-industrial environments yields the dominant contributions to the total dose to the public (Kelly 1987; Jacob and Meckbach 1990). The radiation field is naturally influenced by the environment around the sources. For calculations of the shielding effect of the structures in complex and realistic urban environments Monte Carlo methods turned out to be useful tools (Jacob and Meckbach 1987; Meckbach et al. 1988). Using these methods a complex environment can be set up in which the photon transport can be solved on a reliable way. The accuracy of the methods is in principle limited only by the knowledge of the atomic cross sections and the computational time. Several papers using Monte Carlo results for calculating doses from the external gamma exposures were published (Jacob and Meckbach 1987, 1990; Meckbach et al. 1988; Rochedo et al. 1996). In these papers the Monte Carlo simulations were run in urban environments and for different photon energies. The industrial environment can be defined as such an area where productive and/or commercial activity is carried out. A good example can be a factory or a supermarket. An industrial environment can rather be different from the urban ones as for the types and structures of the buildings and their dimensions. These variations will affect the radiation field of this environment. Hence there is a need to run new Monte Carlo simulations designed specially for the industrial environments

A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom

Energy Technology Data Exchange (ETDEWEB)

Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H., E-mail: mbellezzo@gmail.br [Instituto de Pesquisas Energeticas e Nucleares / CNEN, Av. Lineu Prestes 2242, Cidade Universitaria, 05508-000 Sao Paulo (Brazil)

2014-08-15

As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)

A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom

International Nuclear Information System (INIS)

Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H.

2014-08-01

As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)

Automated Monte Carlo biasing for photon-generated electrons near surfaces.

Energy Technology Data Exchange (ETDEWEB)

Franke, Brian Claude; Crawford, Martin James; Kensek, Ronald Patrick

2009-09-01

This report describes efforts to automate the biasing of coupled electron-photon Monte Carlo particle transport calculations. The approach was based on weight-windows biasing. Weight-window settings were determined using adjoint-flux Monte Carlo calculations. A variety of algorithms were investigated for adaptivity of the Monte Carlo tallies. Tree data structures were used to investigate spatial partitioning. Functional-expansion tallies were used to investigate higher-order spatial representations.

The application of Monte Carlo method to electron and photon beams transport; Zastosowanie metody Monte Carlo do analizy transportu elektronow i fotonow

Energy Technology Data Exchange (ETDEWEB)

Zychor, I. [Soltan Inst. for Nuclear Studies, Otwock-Swierk (Poland)

1994-12-31

The application of a Monte Carlo method to study a transport in matter of electron and photon beams is presented, especially for electrons with energies up to 18 MeV. The SHOWME Monte Carlo code, a modified version of GEANT3 code, was used on the CONVEX C3210 computer at Swierk. It was assumed that an electron beam is mono directional and monoenergetic. Arbitrary user-defined, complex geometries made of any element or material can be used in calculation. All principal phenomena occurring when electron beam penetrates the matter are taken into account. The use of calculation for a therapeutic electron beam collimation is presented. (author). 20 refs, 29 figs.

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

International Nuclear Information System (INIS)

Halbleib, J.A.; Kensek, R.P.; Valdez, G.D.; Mehlhorn, T.A.; Seltzer, S.M.; Berger, M.J.

1993-01-01

ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields. It combines operational simplicity and physical accuracy in order to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Flexibility of construction permits tailoring of the codes to specific applications and extension of code capabilities to more complex applications through simple update procedures

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

Energy Technology Data Exchange (ETDEWEB)

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

1993-06-01

ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields. It combines operational simplicity and physical accuracy in order to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Flexibility of construction permits tailoring of the codes to specific applications and extension of code capabilities to more complex applications through simple update procedures.

A computer code package for Monte Carlo photon-electron transport simulation Comparisons with experimental benchmarks

International Nuclear Information System (INIS)

Popescu, Lucretiu M.

2000-01-01

A computer code package (PTSIM) for particle transport Monte Carlo simulation was developed using object oriented techniques of design and programming. A flexible system for simulation of coupled photon, electron transport, facilitating development of efficient simulation applications, was obtained. For photons: Compton and photo-electric effects, pair production and Rayleigh interactions are simulated, while for electrons, a class II condensed history scheme was considered, in which catastrophic interactions (Moeller electron-electron interaction, bremsstrahlung, etc.) are treated in detail and all other interactions with reduced individual effect on electron history are grouped together using continuous slowing down approximation and energy straggling theories. Electron angular straggling is simulated using Moliere theory or a mixed model in which scatters at large angles are treated as distinct events. Comparisons with experimentally benchmarks for electron transmission and bremsstrahlung emissions energy and angular spectra, and for dose calculations are presented

The MC21 Monte Carlo Transport Code

International Nuclear Information System (INIS)

Sutton TM; Donovan TJ; Trumbull TH; Dobreff PS; Caro E; Griesheimer DP; Tyburski LJ; Carpenter DC; Joo H

2007-01-01

MC21 is a new Monte Carlo neutron and photon transport code currently under joint development at the Knolls Atomic Power Laboratory and the Bettis Atomic Power Laboratory. MC21 is the Monte Carlo transport kernel of the broader Common Monte Carlo Design Tool (CMCDT), which is also currently under development. The vision for CMCDT is to provide an automated, computer-aided modeling and post-processing environment integrated with a Monte Carlo solver that is optimized for reactor analysis. CMCDT represents a strategy to push the Monte Carlo method beyond its traditional role as a benchmarking tool or ''tool of last resort'' and into a dominant design role. This paper describes various aspects of the code, including the neutron physics and nuclear data treatments, the geometry representation, and the tally and depletion capabilities

Accelerating Monte Carlo simulations of photon transport in a voxelized geometry using a massively parallel graphics processing unit

International Nuclear Information System (INIS)

Badal, Andreu; Badano, Aldo

2009-01-01

Purpose: It is a known fact that Monte Carlo simulations of radiation transport are computationally intensive and may require long computing times. The authors introduce a new paradigm for the acceleration of Monte Carlo simulations: The use of a graphics processing unit (GPU) as the main computing device instead of a central processing unit (CPU). Methods: A GPU-based Monte Carlo code that simulates photon transport in a voxelized geometry with the accurate physics models from PENELOPE has been developed using the CUDA programming model (NVIDIA Corporation, Santa Clara, CA). Results: An outline of the new code and a sample x-ray imaging simulation with an anthropomorphic phantom are presented. A remarkable 27-fold speed up factor was obtained using a GPU compared to a single core CPU. Conclusions: The reported results show that GPUs are currently a good alternative to CPUs for the simulation of radiation transport. Since the performance of GPUs is currently increasing at a faster pace than that of CPUs, the advantages of GPU-based software are likely to be more pronounced in the future.

Accelerating Monte Carlo simulations of photon transport in a voxelized geometry using a massively parallel graphics processing unit

Energy Technology Data Exchange (ETDEWEB)

Badal, Andreu; Badano, Aldo [Division of Imaging and Applied Mathematics, OSEL, CDRH, U.S. Food and Drug Administration, Silver Spring, Maryland 20993-0002 (United States)

2009-11-15

Purpose: It is a known fact that Monte Carlo simulations of radiation transport are computationally intensive and may require long computing times. The authors introduce a new paradigm for the acceleration of Monte Carlo simulations: The use of a graphics processing unit (GPU) as the main computing device instead of a central processing unit (CPU). Methods: A GPU-based Monte Carlo code that simulates photon transport in a voxelized geometry with the accurate physics models from PENELOPE has been developed using the CUDA programming model (NVIDIA Corporation, Santa Clara, CA). Results: An outline of the new code and a sample x-ray imaging simulation with an anthropomorphic phantom are presented. A remarkable 27-fold speed up factor was obtained using a GPU compared to a single core CPU. Conclusions: The reported results show that GPUs are currently a good alternative to CPUs for the simulation of radiation transport. Since the performance of GPUs is currently increasing at a faster pace than that of CPUs, the advantages of GPU-based software are likely to be more pronounced in the future.

Accelerating Monte Carlo simulations of photon transport in a voxelized geometry using a massively parallel graphics processing unit.

Science.gov (United States)

Badal, Andreu; Badano, Aldo

2009-11-01

It is a known fact that Monte Carlo simulations of radiation transport are computationally intensive and may require long computing times. The authors introduce a new paradigm for the acceleration of Monte Carlo simulations: The use of a graphics processing unit (GPU) as the main computing device instead of a central processing unit (CPU). A GPU-based Monte Carlo code that simulates photon transport in a voxelized geometry with the accurate physics models from PENELOPE has been developed using the CUDATM programming model (NVIDIA Corporation, Santa Clara, CA). An outline of the new code and a sample x-ray imaging simulation with an anthropomorphic phantom are presented. A remarkable 27-fold speed up factor was obtained using a GPU compared to a single core CPU. The reported results show that GPUs are currently a good alternative to CPUs for the simulation of radiation transport. Since the performance of GPUs is currently increasing at a faster pace than that of CPUs, the advantages of GPU-based software are likely to be more pronounced in the future.

Statistics of Monte Carlo methods used in radiation transport calculation

International Nuclear Information System (INIS)

Datta, D.

2009-01-01

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

Space applications of the MITS electron-photon Monte Carlo transport code system

International Nuclear Information System (INIS)

Kensek, R.P.; Lorence, L.J.; Halbleib, J.A.; Morel, J.E.

1996-01-01

The MITS multigroup/continuous-energy electron-photon Monte Carlo transport code system has matured to the point that it is capable of addressing more realistic three-dimensional adjoint applications. It is first employed to efficiently predict point doses as a function of source energy for simple three-dimensional experimental geometries exposed to simulated uniform isotropic planar sources of monoenergetic electrons up to 4.0 MeV. Results are in very good agreement with experimental data. It is then used to efficiently simulate dose to a detector in a subsystem of a GPS satellite due to its natural electron environment, employing a relatively complex model of the satellite. The capability for survivability analysis of space systems is demonstrated, and results are obtained with and without variance reduction

Monte Carlo photon transport on shared memory and distributed memory parallel processors

International Nuclear Information System (INIS)

Martin, W.R.; Wan, T.C.; Abdel-Rahman, T.S.; Mudge, T.N.; Miura, K.

1987-01-01

Parallelized Monte Carlo algorithms for analyzing photon transport in an inertially confined fusion (ICF) plasma are considered. Algorithms were developed for shared memory (vector and scalar) and distributed memory (scalar) parallel processors. The shared memory algorithm was implemented on the IBM 3090/400, and timing results are presented for dedicated runs with two, three, and four processors. Two alternative distributed memory algorithms (replication and dispatching) were implemented on a hypercube parallel processor (1 through 64 nodes). The replication algorithm yields essentially full efficiency for all cube sizes; with the 64-node configuration, the absolute performance is nearly the same as with the CRAY X-MP. The dispatching algorithm also yields efficiencies above 80% in a large simulation for the 64-processor configuration

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

International Nuclear Information System (INIS)

Deng Li; Xie Zhongsheng

1999-01-01

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

Simulation of photon and charge transport in X-ray imaging semiconductor sensors

CERN Document Server

Nilsson, H E; Hjelm, M; Bertilsson, K

2002-01-01

A fully stochastic model for the imaging properties of X-ray silicon pixel detectors is presented. Both integrating and photon counting configurations have been considered, as well as scintillator-coated structures. The model is based on three levels of Monte Carlo simulations; photon transport and absorption using MCNP, full band Monte Carlo simulation of charge transport and system level Monte Carlo simulation of the imaging performance of the detector system. In the case of scintillator-coated detectors, the light scattering in the detector layers has been simulated using a Monte Carlo method. The image resolution was found to be much lower in scintillator-coated systems due to large light spread in thick scintillator layers. A comparison between integrating and photon counting readout methods shows that the image resolution can be slightly enhanced using a photon-counting readout. In addition, the proposed model has been used to study charge-sharing effects on the energy resolution in photon counting dete...

Scalable and massively parallel Monte Carlo photon transport simulations for heterogeneous computing platforms.

Science.gov (United States)

Yu, Leiming; Nina-Paravecino, Fanny; Kaeli, David; Fang, Qianqian

2018-01-01

We present a highly scalable Monte Carlo (MC) three-dimensional photon transport simulation platform designed for heterogeneous computing systems. Through the development of a massively parallel MC algorithm using the Open Computing Language framework, this research extends our existing graphics processing unit (GPU)-accelerated MC technique to a highly scalable vendor-independent heterogeneous computing environment, achieving significantly improved performance and software portability. A number of parallel computing techniques are investigated to achieve portable performance over a wide range of computing hardware. Furthermore, multiple thread-level and device-level load-balancing strategies are developed to obtain efficient simulations using multiple central processing units and GPUs. (2018) COPYRIGHT Society of Photo-Optical Instrumentation Engineers (SPIE).

SU-E-T-558: Monte Carlo Photon Transport Simulations On GPU with Quadric Geometry

International Nuclear Information System (INIS)

Chi, Y; Tian, Z; Jiang, S; Jia, X

2015-01-01

Purpose: Monte Carlo simulation on GPU has experienced rapid advancements over the past a few years and tremendous accelerations have been achieved. Yet existing packages were developed only in voxelized geometry. In some applications, e.g. radioactive seed modeling, simulations in more complicated geometry are needed. This abstract reports our initial efforts towards developing a quadric geometry module aiming at expanding the application scope of GPU-based MC simulations. Methods: We defined the simulation geometry consisting of a number of homogeneous bodies, each specified by its material composition and limiting surfaces characterized by quadric functions. A tree data structure was utilized to define geometric relationship between different bodies. We modified our GPU-based photon MC transport package to incorporate this geometry. Specifically, geometry parameters were loaded into GPU’s shared memory for fast access. Geometry functions were rewritten to enable the identification of the body that contains the current particle location via a fast searching algorithm based on the tree data structure. Results: We tested our package in an example problem of HDR-brachytherapy dose calculation for shielded cylinder. The dose under the quadric geometry and that under the voxelized geometry agreed in 94.2% of total voxels within 20% isodose line based on a statistical t-test (95% confidence level), where the reference dose was defined to be the one at 0.5cm away from the cylinder surface. It took 243sec to transport 100million source photons under this quadric geometry on an NVidia Titan GPU card. Compared with simulation time of 99.6sec in the voxelized geometry, including quadric geometry reduced efficiency due to the complicated geometry-related computations. Conclusion: Our GPU-based MC package has been extended to support photon transport simulation in quadric geometry. Satisfactory accuracy was observed with a reduced efficiency. Developments for charged

SU-E-T-558: Monte Carlo Photon Transport Simulations On GPU with Quadric Geometry

Energy Technology Data Exchange (ETDEWEB)

Chi, Y; Tian, Z; Jiang, S; Jia, X [The University of Texas Southwestern Medical Ctr, Dallas, TX (United States)

2015-06-15

Purpose: Monte Carlo simulation on GPU has experienced rapid advancements over the past a few years and tremendous accelerations have been achieved. Yet existing packages were developed only in voxelized geometry. In some applications, e.g. radioactive seed modeling, simulations in more complicated geometry are needed. This abstract reports our initial efforts towards developing a quadric geometry module aiming at expanding the application scope of GPU-based MC simulations. Methods: We defined the simulation geometry consisting of a number of homogeneous bodies, each specified by its material composition and limiting surfaces characterized by quadric functions. A tree data structure was utilized to define geometric relationship between different bodies. We modified our GPU-based photon MC transport package to incorporate this geometry. Specifically, geometry parameters were loaded into GPU’s shared memory for fast access. Geometry functions were rewritten to enable the identification of the body that contains the current particle location via a fast searching algorithm based on the tree data structure. Results: We tested our package in an example problem of HDR-brachytherapy dose calculation for shielded cylinder. The dose under the quadric geometry and that under the voxelized geometry agreed in 94.2% of total voxels within 20% isodose line based on a statistical t-test (95% confidence level), where the reference dose was defined to be the one at 0.5cm away from the cylinder surface. It took 243sec to transport 100million source photons under this quadric geometry on an NVidia Titan GPU card. Compared with simulation time of 99.6sec in the voxelized geometry, including quadric geometry reduced efficiency due to the complicated geometry-related computations. Conclusion: Our GPU-based MC package has been extended to support photon transport simulation in quadric geometry. Satisfactory accuracy was observed with a reduced efficiency. Developments for charged

MCPT: A Monte Carlo code for simulation of photon transport in tomographic scanners

International Nuclear Information System (INIS)

Prettyman, T.H.; Gardner, R.P.; Verghese, K.

1990-01-01

MCPT is a special-purpose Monte Carlo code designed to simulate photon transport in tomographic scanners. Variance reduction schemes and sampling games present in MCPT were selected to characterize features common to most tomographic scanners. Combined splitting and biasing (CSB) games are used to systematically sample important detection pathways. An efficient splitting game is used to tally particle energy deposition in detection zones. The pulse height distribution of each detector can be found by convolving the calculated energy deposition distribution with the detector's resolution function. A general geometric modelling package, HERMETOR, is used to describe the geometry of the tomographic scanners and provide MCPT information needed for particle tracking. MCPT's modelling capabilites are described and preliminary experimental validation is presented. (orig.)

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

Optix: A Monte Carlo scintillation light transport code

Energy Technology Data Exchange (ETDEWEB)

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

2014-02-11

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

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

International Nuclear Information System (INIS)

1978-07-01

The general-purpose Monte Carlo code MCNP can be used for neutron, photon, or coupled neutron--photon transport. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-IV) are accounted for. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. Standard optional variance reduction schemes include geometry splitting and Russian roulette, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point detectors, track-length estimators, and source biasing. The standard output of MCNP includes two-way current as a function of energy, time, and angle with the normal, across any subset of bounding surfaces in the problem. Fluxes across any set of bounding surfaces are available as a function of time and energy. Similarly, the flux at designated points and the average flux in a cell (track length per unit volume) are standard tallies. Reactions such as fissions or absorptions may be obtained in a subset of geometric cells. The heating tallies give the energy deposition per starting particle. In addition, particles may be flagged when they cross specified surfaces or enter designated cells, and the contributions of these flagged particles to certain of the tallies are listed separately. All quantities printed out have their relative errors listed also. 11 figures, 27 tables

Application of the Monte Carlo method in calculation of energy-time distribution from a pulsed photon source in homogeneous air environment

International Nuclear Information System (INIS)

Ilic, R.D.; Vojvodic, V.I.; Orlic, M.P.

1981-01-01

The stochastic nature of photon interactions with matter and the characteristics of photon transport through real materials, are very well suited for applications of the Monte Carlo method in calculations of the energy-space distribution of photons. Starting from general principles of the Monte Carlo method, physical-mathematical model of photon transport from a pulsed source is given for the homogeneous air environment. Based on that model, a computer program is written which is applied in calculations of scattered photons delay spectra and changes of the photon energy spectrum. Obtained results provide the estimation of the timespace function of the electromagnetic field generated by photon from a pulsed source. (author)

Effective use of Monte Carlo methods for simulating photon transport with special reference to slab penetration problems in X-ray diagnostics

International Nuclear Information System (INIS)

Carlsson, G.A.

1981-01-01

The analysis of Monte Carlo methods has been made in connection with a particular problem concerning the transport of low energy photons (30-140 keV) through layers of water with thicknesses between 5 and 20 cm. While not claiming to be a complete exposition of available Monte Carlo techniques, the methodological analyses are not restricted to this particular problem. The report describes in a general manner a number of methods which can be used in order to obtain results of greater precision in a fixed computing time. (Auth.)

Ant colony algorithm implementation in electron and photon Monte Carlo transport: application to the commissioning of radiosurgery photon beams.

Science.gov (United States)

García-Pareja, S; Galán, P; Manzano, F; Brualla, L; Lallena, A M

2010-07-01

In this work, the authors describe an approach which has been developed to drive the application of different variance-reduction techniques to the Monte Carlo simulation of photon and electron transport in clinical accelerators. The new approach considers the following techniques: Russian roulette, splitting, a modified version of the directional bremsstrahlung splitting, and the azimuthal particle redistribution. Their application is controlled by an ant colony algorithm based on an importance map. The procedure has been applied to radiosurgery beams. Specifically, the authors have calculated depth-dose profiles, off-axis ratios, and output factors, quantities usually considered in the commissioning of these beams. The agreement between Monte Carlo results and the corresponding measurements is within approximately 3%/0.3 mm for the central axis percentage depth dose and the dose profiles. The importance map generated in the calculation can be used to discuss simulation details in the different parts of the geometry in a simple way. The simulation CPU times are comparable to those needed within other approaches common in this field. The new approach is competitive with those previously used in this kind of problems (PSF generation or source models) and has some practical advantages that make it to be a good tool to simulate the radiation transport in problems where the quantities of interest are difficult to obtain because of low statistics.

Ant colony algorithm implementation in electron and photon Monte Carlo transport: Application to the commissioning of radiosurgery photon beams

Energy Technology Data Exchange (ETDEWEB)

Garcia-Pareja, S.; Galan, P.; Manzano, F.; Brualla, L.; Lallena, A. M. [Servicio de Radiofisica Hospitalaria, Hospital Regional Universitario ' ' Carlos Haya' ' , Avda. Carlos Haya s/n, E-29010 Malaga (Spain); Unidad de Radiofisica Hospitalaria, Hospital Xanit Internacional, Avda. de los Argonautas s/n, E-29630 Benalmadena (Malaga) (Spain); NCTeam, Strahlenklinik, Universitaetsklinikum Essen, Hufelandstr. 55, D-45122 Essen (Germany); Departamento de Fisica Atomica, Molecular y Nuclear, Universidad de Granada, E-18071 Granada (Spain)

2010-07-15

Purpose: In this work, the authors describe an approach which has been developed to drive the application of different variance-reduction techniques to the Monte Carlo simulation of photon and electron transport in clinical accelerators. Methods: The new approach considers the following techniques: Russian roulette, splitting, a modified version of the directional bremsstrahlung splitting, and the azimuthal particle redistribution. Their application is controlled by an ant colony algorithm based on an importance map. Results: The procedure has been applied to radiosurgery beams. Specifically, the authors have calculated depth-dose profiles, off-axis ratios, and output factors, quantities usually considered in the commissioning of these beams. The agreement between Monte Carlo results and the corresponding measurements is within {approx}3%/0.3 mm for the central axis percentage depth dose and the dose profiles. The importance map generated in the calculation can be used to discuss simulation details in the different parts of the geometry in a simple way. The simulation CPU times are comparable to those needed within other approaches common in this field. Conclusions: The new approach is competitive with those previously used in this kind of problems (PSF generation or source models) and has some practical advantages that make it to be a good tool to simulate the radiation transport in problems where the quantities of interest are difficult to obtain because of low statistics.

Ant colony algorithm implementation in electron and photon Monte Carlo transport: Application to the commissioning of radiosurgery photon beams

International Nuclear Information System (INIS)

Garcia-Pareja, S.; Galan, P.; Manzano, F.; Brualla, L.; Lallena, A. M.

2010-01-01

Purpose: In this work, the authors describe an approach which has been developed to drive the application of different variance-reduction techniques to the Monte Carlo simulation of photon and electron transport in clinical accelerators. Methods: The new approach considers the following techniques: Russian roulette, splitting, a modified version of the directional bremsstrahlung splitting, and the azimuthal particle redistribution. Their application is controlled by an ant colony algorithm based on an importance map. Results: The procedure has been applied to radiosurgery beams. Specifically, the authors have calculated depth-dose profiles, off-axis ratios, and output factors, quantities usually considered in the commissioning of these beams. The agreement between Monte Carlo results and the corresponding measurements is within ∼3%/0.3 mm for the central axis percentage depth dose and the dose profiles. The importance map generated in the calculation can be used to discuss simulation details in the different parts of the geometry in a simple way. The simulation CPU times are comparable to those needed within other approaches common in this field. Conclusions: The new approach is competitive with those previously used in this kind of problems (PSF generation or source models) and has some practical advantages that make it to be a good tool to simulate the radiation transport in problems where the quantities of interest are difficult to obtain because of low statistics.

ITS Version 6 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes.

Energy Technology Data Exchange (ETDEWEB)

Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William

2008-04-01

ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of lineartime-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 90. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-07-01

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

Advanced Monte Carlo for radiation physics, particle transport simulation and applications. Proceedings

International Nuclear Information System (INIS)

Kling, A.; Barao, F.J.C.; Nakagawa, M.; Tavora, L.

2001-01-01

The following topics were dealt with: Electron and photon interactions and transport mechanisms, random number generation, applications in medical physisc, microdosimetry, track structure, radiobiological modeling, Monte Carlo method in radiotherapy, dosimetry, and medical accelerator simulation, neutron transport, high-energy hadron transport. (HSI)

Implementation of a Monte Carlo method to model photon conversion for solar cells

International Nuclear Information System (INIS)

Canizo, C. del; Tobias, I.; Perez-Bedmar, J.; Pan, A.C.; Luque, A.

2008-01-01

A physical model describing different photon conversion mechanisms is presented in the context of photovoltaic applications. To solve the resulting system of equations, a Monte Carlo ray-tracing model is implemented, which takes into account the coupling of the photon transport phenomena to the non-linear rate equations describing luminescence. It also separates the generation of rays from the two very different sources of photons involved (the sun and the luminescence centers). The Monte Carlo simulator presented in this paper is proposed as a tool to help in the evaluation of candidate materials for up- and down-conversion. Some application examples are presented, exploring the range of values that the most relevant parameters describing the converter should have in order to give significant gain in photocurrent

Comparative evaluation of photon cross section libraries for materials of interest in PET Monte Carlo simulations

CERN Document Server

Zaidi, H

1999-01-01

the many applications of Monte Carlo modelling in nuclear medicine imaging make it desirable to increase the accuracy and computational speed of Monte Carlo codes. The accuracy of Monte Carlo simulations strongly depends on the accuracy in the probability functions and thus on the cross section libraries used for photon transport calculations. A comparison between different photon cross section libraries and parametrizations implemented in Monte Carlo simulation packages developed for positron emission tomography and the most recent Evaluated Photon Data Library (EPDL97) developed by the Lawrence Livermore National Laboratory was performed for several human tissues and common detector materials for energies from 1 keV to 1 MeV. Different photon cross section libraries and parametrizations show quite large variations as compared to the EPDL97 coefficients. This latter library is more accurate and was carefully designed in the form of look-up tables providing efficient data storage, access, and management. Toge...

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

International Nuclear Information System (INIS)

Cullen, D.E

2000-01-01

TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files

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

CERN Document Server

Cullen, D

2000-01-01

TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files.

penORNL: a parallel Monte Carlo photon and electron transport package using PENELOPE

International Nuclear Information System (INIS)

Bekar, Kursat B.; Miller, Thomas Martin; Patton, Bruce W.; Weber, Charles F.

2015-01-01

The parallel Monte Carlo photon and electron transport code package penORNL was developed at Oak Ridge National Laboratory to enable advanced scanning electron microscope (SEM) simulations on high-performance computing systems. This paper discusses the implementations, capabilities and parallel performance of the new code package. penORNL uses PENELOPE for its physics calculations and provides all available PENELOPE features to the users, as well as some new features including source definitions specifically developed for SEM simulations, a pulse-height tally capability for detailed simulations of gamma and x-ray detectors, and a modified interaction forcing mechanism to enable accurate energy deposition calculations. The parallel performance of penORNL was extensively tested with several model problems, and very good linear parallel scaling was observed with up to 512 processors. penORNL, along with its new features, will be available for SEM simulations upon completion of the new pulse-height tally implementation.

Modelling of neutron and photon transport in iron and concrete radiation shieldings by the Monte Carlo method - Version 2

CERN Document Server

Žukauskaite, A; Plukiene, R; Plukis, A

2007-01-01

Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 – γ-ray beams (1-10 MeV), HIMAC and ISIS-800 – high energy neutrons (20-800 MeV) transport in iron and concrete. The results were then compared with experimental data.

Correlated Production and Analog Transport of Fission Neutrons and Photons using Fission Models FREYA, FIFRELIN and the Monte Carlo Code TRIPOLI-4® .

Science.gov (United States)

Verbeke, Jérôme M.; Petit, Odile; Chebboubi, Abdelhazize; Litaize, Olivier

2018-01-01

Fission modeling in general-purpose Monte Carlo transport codes often relies on average nuclear data provided by international evaluation libraries. As such, only average fission multiplicities are available and correlations between fission neutrons and photons are missing. Whereas uncorrelated fission physics is usually sufficient for standard reactor core and radiation shielding calculations, correlated fission secondaries are required for specialized nuclear instrumentation and detector modeling. For coincidence counting detector optimization for instance, precise simulation of fission neutrons and photons that remain correlated in time from birth to detection is essential. New developments were recently integrated into the Monte Carlo transport code TRIPOLI-4 to model fission physics more precisely, the purpose being to access event-by-event fission events from two different fission models: FREYA and FIFRELIN. TRIPOLI-4 simulations can now be performed, either by connecting via an API to the LLNL fission library including FREYA, or by reading external fission event data files produced by FIFRELIN beforehand. These new capabilities enable us to easily compare results from Monte Carlo transport calculations using the two fission models in a nuclear instrumentation application. In the first part of this paper, broad underlying principles of the two fission models are recalled. We then present experimental measurements of neutron angular correlations for 252Cf(sf) and 240Pu(sf). The correlations were measured for several neutron kinetic energy thresholds. In the latter part of the paper, simulation results are compared to experimental data. Spontaneous fissions in 252Cf and 240Pu are modeled by FREYA or FIFRELIN. Emitted neutrons and photons are subsequently transported to an array of scintillators by TRIPOLI-4 in analog mode to preserve their correlations. Angular correlations between fission neutrons obtained independently from these TRIPOLI-4 simulations, using

Coupled electron/photon transport in static external magnetic fields

International Nuclear Information System (INIS)

Halbleib, J.A. Sr.; Vandevender, W.H.

A model is presented which describes coupled electron/photon transport in the presence of static magnetic fields of arbitrary spatial dependence. The method combines state-of-the-art condensed-history electron collisional Monte Carlo and single-scattering photon Monte Carlo, including electron energy-loss straggling and the production and transport of all generations of secondaries, with numerical field integration via the best available variable-step-size Runge-Kutta-Fehlberg or variable-order/variable-step-size Adams PECE differential equation solvers. A three-dimensional cartesian system is employed in the description of particle trajectories. Although the present model is limited to multilayer material configurations, extension to more complex material geometries should not be difficult. Among the more important options are (1) a feature which permits the neglect of field effects in regions where transport is collision dominated and (2) a method for describing the transport in variable-density media where electron energies and material densities are sufficiently low that the density effect on electronic stopping powers may be neglected. (U.S.)

PENELOPE, and algorithm and computer code for Monte Carlo simulation of electron-photon showers

Energy Technology Data Exchange (ETDEWEB)

Salvat, F.; Fernandez-Varea, J.M.; Baro, J.; Sempau, J.

1996-10-01

The FORTRAN 77 subroutine package PENELOPE performs Monte Carlo simulation of electron-photon showers in arbitrary for a wide energy range, from similar{sub t}o 1 KeV to several hundred MeV. Photon transport is simulated by means of the standard, detailed simulation scheme. Electron and positron histories are generated on the basis of a mixed procedure, which combines detailed simulation of hard events with condensed simulation of soft interactions. A simple geometry package permits the generation of random electron-photon showers in material systems consisting of homogeneous bodies limited by quadric surfaces, i.e. planes, spheres cylinders, etc. This report is intended not only to serve as a manual of the simulation package, but also to provide the user with the necessary information to understand the details of the Monte Carlo algorithm.

PENELOPE, an algorithm and computer code for Monte Carlo simulation of electron-photon showers

Energy Technology Data Exchange (ETDEWEB)

Salvat, F; Fernandez-Varea, J M; Baro, J; Sempau, J

1996-07-01

The FORTRAN 77 subroutine package PENELOPE performs Monte Carlo simulation of electron-photon showers in arbitrary for a wide energy range, from 1 keV to several hundred MeV. Photon transport is simulated by means of the standard, detailed simulation scheme. Electron and positron histories are generated on the basis of a mixed procedure, which combines detailed simulation of hard events with condensed simulation of soft interactions. A simple geometry package permits the generation of random electron-photon showers in material systems consisting of homogeneous bodies limited by quadric surfaces, i.e. planes, spheres, cylinders, etc. This report is intended not only to serve as a manual of the simulation package, but also to provide the user with the necessary information to understand the details of the Monte Carlo algorithm. (Author) 108 refs.

TRIPOLI-4: Monte Carlo transport code functionalities and applications

International Nuclear Information System (INIS)

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

2003-01-01

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

Advantages of Analytical Transformations in Monte Carlo Methods for Radiation Transport

International Nuclear Information System (INIS)

McKinley, M S; Brooks III, E D; Daffin, F

2004-01-01

Monte Carlo methods for radiation transport typically attempt to solve an integral by directly sampling analog or weighted particles, which are treated as physical entities. Improvements to the methods involve better sampling, probability games or physical intuition about the problem. We show that significant improvements can be achieved by recasting the equations with an analytical transform to solve for new, non-physical entities or fields. This paper looks at one such transform, the difference formulation for thermal photon transport, showing a significant advantage for Monte Carlo solution of the equations for time dependent transport. Other related areas are discussed that may also realize significant benefits from similar analytical transformations

DEEP code to calculate dose equivalents in human phantom for external photon exposure by Monte Carlo method

International Nuclear Information System (INIS)

Yamaguchi, Yasuhiro

1991-01-01

The present report describes a computer code DEEP which calculates the organ dose equivalents and the effective dose equivalent for external photon exposure by the Monte Carlo method. MORSE-CG, Monte Carlo radiation transport code, is incorporated into the DEEP code to simulate photon transport phenomena in and around a human body. The code treats an anthropomorphic phantom represented by mathematical formulae and user has a choice for the phantom sex: male, female and unisex. The phantom can wear personal dosimeters on it and user can specify their location and dimension. This document includes instruction and sample problem for the code as well as the general description of dose calculation, human phantom and computer code. (author)

A method for photon beam Monte Carlo multileaf collimator particle transport

Science.gov (United States)

Siebers, Jeffrey V.; Keall, Paul J.; Kim, Jong Oh; Mohan, Radhe

2002-09-01

Monte Carlo (MC) algorithms are recognized as the most accurate methodology for patient dose assessment. For intensity-modulated radiation therapy (IMRT) delivered with dynamic multileaf collimators (DMLCs), accurate dose calculation, even with MC, is challenging. Accurate IMRT MC dose calculations require inclusion of the moving MLC in the MC simulation. Due to its complex geometry, full transport through the MLC can be time consuming. The aim of this work was to develop an MLC model for photon beam MC IMRT dose computations. The basis of the MC MLC model is that the complex MLC geometry can be separated into simple geometric regions, each of which readily lends itself to simplified radiation transport. For photons, only attenuation and first Compton scatter interactions are considered. The amount of attenuation material an individual particle encounters while traversing the entire MLC is determined by adding the individual amounts from each of the simplified geometric regions. Compton scatter is sampled based upon the total thickness traversed. Pair production and electron interactions (scattering and bremsstrahlung) within the MLC are ignored. The MLC model was tested for 6 MV and 18 MV photon beams by comparing it with measurements and MC simulations that incorporate the full physics and geometry for fields blocked by the MLC and with measurements for fields with the maximum possible tongue-and-groove and tongue-or-groove effects, for static test cases and for sliding windows of various widths. The MLC model predicts the field size dependence of the MLC leakage radiation within 0.1% of the open-field dose. The entrance dose and beam hardening behind a closed MLC are predicted within +/-1% or 1 mm. Dose undulations due to differences in inter- and intra-leaf leakage are also correctly predicted. The MC MLC model predicts leaf-edge tongue-and-groove dose effect within +/-1% or 1 mm for 95% of the points compared at 6 MV and 88% of the points compared at 18 MV

A method for photon beam Monte Carlo multileaf collimator particle transport

Energy Technology Data Exchange (ETDEWEB)

Siebers, Jeffrey V. [Department of Radiation Oncology, Medical College of Virginia Hospitals, Virginia Commonwealth University, Richmond, VA (United States)]. E-mail: jsiebers@vcu.edu; Keall, Paul J.; Kim, Jong Oh; Mohan, Radhe [Department of Radiation Oncology, Medical College of Virginia Hospitals, Virginia Commonwealth University, Richmond, VA (United States)

2002-09-07

Monte Carlo (MC) algorithms are recognized as the most accurate methodology for patient dose assessment. For intensity-modulated radiation therapy (IMRT) delivered with dynamic multileaf collimators (DMLCs), accurate dose calculation, even with MC, is challenging. Accurate IMRT MC dose calculations require inclusion of the moving MLC in the MC simulation. Due to its complex geometry, full transport through the MLC can be time consuming. The aim of this work was to develop an MLC model for photon beam MC IMRT dose computations. The basis of the MC MLC model is that the complex MLC geometry can be separated into simple geometric regions, each of which readily lends itself to simplified radiation transport. For photons, only attenuation and first Compton scatter interactions are considered. The amount of attenuation material an individual particle encounters while traversing the entire MLC is determined by adding the individual amounts from each of the simplified geometric regions. Compton scatter is sampled based upon the total thickness traversed. Pair production and electron interactions (scattering and bremsstrahlung) within the MLC are ignored. The MLC model was tested for 6 MV and 18 MV photon beams by comparing it with measurements and MC simulations that incorporate the full physics and geometry for fields blocked by the MLC and with measurements for fields with the maximum possible tongue-and-groove and tongue-or-groove effects, for static test cases and for sliding windows of various widths. The MLC model predicts the field size dependence of the MLC leakage radiation within 0.1% of the open-field dose. The entrance dose and beam hardening behind a closed MLC are predicted within {+-}1% or 1 mm. Dose undulations due to differences in inter- and intra-leaf leakage are also correctly predicted. The MC MLC model predicts leaf-edge tongue-and-groove dose effect within {+-}1% or 1 mm for 95% of the points compared at 6 MV and 88% of the points compared at 18 MV

Optimization of the Monte Carlo code for modeling of photon migration in tissue.

Science.gov (United States)

Zołek, Norbert S; Liebert, Adam; Maniewski, Roman

2006-10-01

The Monte Carlo method is frequently used to simulate light transport in turbid media because of its simplicity and flexibility, allowing to analyze complicated geometrical structures. Monte Carlo simulations are, however, time consuming because of the necessity to track the paths of individual photons. The time consuming computation is mainly associated with the calculation of the logarithmic and trigonometric functions as well as the generation of pseudo-random numbers. In this paper, the Monte Carlo algorithm was developed and optimized, by approximation of the logarithmic and trigonometric functions. The approximations were based on polynomial and rational functions, and the errors of these approximations are less than 1% of the values of the original functions. The proposed algorithm was verified by simulations of the time-resolved reflectance at several source-detector separations. The results of the calculation using the approximated algorithm were compared with those of the Monte Carlo simulations obtained with an exact computation of the logarithm and trigonometric functions as well as with the solution of the diffusion equation. The errors of the moments of the simulated distributions of times of flight of photons (total number of photons, mean time of flight and variance) are less than 2% for a range of optical properties, typical of living tissues. The proposed approximated algorithm allows to speed up the Monte Carlo simulations by a factor of 4. The developed code can be used on parallel machines, allowing for further acceleration.

Direct utilization of information from nuclear data files in Monte Carlo simulation of neutron and photon transport

International Nuclear Information System (INIS)

Androsenko, P.; Joloudov, D.; Kompaniyets, A.

2001-01-01

Questions, related to Monte-Carlo method for solution of neutron and photon transport equation, are discussed in the work concerned. Problems dealing with direct utilization of information from evaluated nuclear data files in run-time calculations are considered. ENDF-6 format libraries have been used for calculations. Approaches provided by the rules of ENDF-6 files 2, 3-6, 12-15, 23, 27 and algorithms for reconstruction of resolved and unresolved resonance region cross sections under preset energy are described. The comparison results of calculations made by NJOY and GRUCON programs and computed cross sections data are represented. Test computation data of neutron leakage spectra for spherical benchmark-experiments are also represented. (authors)

Modeling of neutron and photon transport in iron and concrete radiation shields by using Monte Carlo method

CERN Document Server

Žukauskaitėa, A; Plukienė, R; Ridikas, D

2007-01-01

Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 (AVF cyclotron of Research Center of Nuclear Physics, Osaka University, Japan) – γ-ray beams (1-10 MeV), HIMAC (heavy-ion synchrotron of the National Institute of Radiological Sciences in Chiba, Japan) and ISIS-800 (ISIS intensive spallation neutron source facility of the Rutherford Appleton laboratory, UK) – high energy neutron (20-800 MeV) transport in iron and concrete. The calculation results were then compared with experimental data.compared with experimental data.

A photon source model based on particle transport in a parameterized accelerator structure for Monte Carlo dose calculations.

Science.gov (United States)

Ishizawa, Yoshiki; Dobashi, Suguru; Kadoya, Noriyuki; Ito, Kengo; Chiba, Takahito; Takayama, Yoshiki; Sato, Kiyokazu; Takeda, Ken

2018-05-17

An accurate source model of a medical linear accelerator is essential for Monte Carlo (MC) dose calculations. This study aims to propose an analytical photon source model based on particle transport in parameterized accelerator structures, focusing on a more realistic determination of linac photon spectra compared to existing approaches. We designed the primary and secondary photon sources based on the photons attenuated and scattered by a parameterized flattening filter. The primary photons were derived by attenuating bremsstrahlung photons based on the path length in the filter. Conversely, the secondary photons were derived from the decrement of the primary photons in the attenuation process. This design facilitates these sources to share the free parameters of the filter shape and be related to each other through the photon interaction in the filter. We introduced two other parameters of the primary photon source to describe the particle fluence in penumbral regions. All the parameters are optimized based on calculated dose curves in water using the pencil-beam-based algorithm. To verify the modeling accuracy, we compared the proposed model with the phase space data (PSD) of the Varian TrueBeam 6 and 15 MV accelerators in terms of the beam characteristics and the dose distributions. The EGS5 Monte Carlo code was used to calculate the dose distributions associated with the optimized model and reference PSD in a homogeneous water phantom and a heterogeneous lung phantom. We calculated the percentage of points passing 1D and 2D gamma analysis with 1%/1 mm criteria for the dose curves and lateral dose distributions, respectively. The optimized model accurately reproduced the spectral curves of the reference PSD both on- and off-axis. The depth dose and lateral dose profiles of the optimized model also showed good agreement with those of the reference PSD. The passing rates of the 1D gamma analysis with 1%/1 mm criteria between the model and PSD were 100% for 4�

Coupled electron-photon radiation transport

International Nuclear Information System (INIS)

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

2000-01-01

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

Monte Carlo Transport for Electron Thermal Transport

Science.gov (United States)

Chenhall, Jeffrey; Cao, Duc; Moses, Gregory

2015-11-01

The iSNB (implicit Schurtz Nicolai Busquet multigroup electron thermal transport method of Cao et al. is adapted into a Monte Carlo transport method in order to better model the effects of non-local behavior. The end goal is a hybrid transport-diffusion method that combines Monte Carlo Transport with a discrete diffusion Monte Carlo (DDMC). The hybrid method will combine the efficiency of a diffusion method in short mean free path regions with the accuracy of a transport method in long mean free path regions. The Monte Carlo nature of the approach allows the algorithm to be massively parallelized. Work to date on the method will be presented. This work was supported by Sandia National Laboratory - Albuquerque and the University of Rochester Laboratory for Laser Energetics.

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

International Nuclear Information System (INIS)

Shangguan Danhua; Li Gang; Deng Li; Zhang Baoyin

2013-01-01

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

Monte Carlo electron-photon transport using GPUs as an accelerator: Results for a water-aluminum-water phantom

Energy Technology Data Exchange (ETDEWEB)

Su, L.; Du, X.; Liu, T.; Xu, X. G. [Nuclear Engineering Program, Rensselaer Polytechnic Institute, Troy, NY 12180 (United States)

2013-07-01

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. In this paper, the preliminary results of code development and testing are presented. The electron transport in media was modeled using the class-II condensed history method. The electron energy considered ranges from a few hundred keV to 30 MeV. Moller scattering and bremsstrahlung processes above a preset energy were explicitly modeled. Energy loss below that threshold was accounted for using the Continuously Slowing Down Approximation (CSDA). Photon transport was dealt with using the delta tracking method. Photoelectric effect, Compton scattering and pair production were modeled. Voxelised geometry was supported. A serial ARHCHER-CPU was first written in C++. The code was then ported to the GPU platform using CUDA C. The hardware involved a desktop PC with an Intel Xeon X5660 CPU and six NVIDIA Tesla M2090 GPUs. ARHCHER was tested for a case of 20 MeV electron beam incident perpendicularly on a water-aluminum-water phantom. The depth and lateral dose profiles were found to agree with results obtained from well tested MC codes. Using six GPU cards, 6x10{sup 6} histories of electrons were simulated within 2 seconds. In comparison, the same case running the EGSnrc and MCNPX codes required 1645 seconds and 9213 seconds, respectively, on a CPU with a single core used. (authors)

Tripoli-4, a three-dimensional poly-kinetic particle transport Monte-Carlo code

International Nuclear Information System (INIS)

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

2003-01-01

In this updated of the Monte-Carlo transport code Tripoli-4, we list and describe its current main features. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. Tripoli-4 enables the user to compute the following physical quantities: a flux, a multiplication factor, a current, a reaction rate, a dose equivalent rate as well as deposit of energy and recoil energies. For each interesting physical quantity, a Monte-Carlo simulation offers different types of estimators. Tripoli-4 has support for execution in parallel mode. Special features and applications are also presented

Tripoli-4, a three-dimensional poly-kinetic particle transport Monte-Carlo code

Energy Technology Data Exchange (ETDEWEB)

Both, J P; Lee, Y K; Mazzolo, A; Peneliau, Y; Petit, O; Roesslinger, B; Soldevila, M [CEA Saclay, Dir. de l' Energie Nucleaire (DEN/DM2S/SERMA/LEPP), 91 - Gif sur Yvette (France)

2003-07-01

In this updated of the Monte-Carlo transport code Tripoli-4, we list and describe its current main features. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. Tripoli-4 enables the user to compute the following physical quantities: a flux, a multiplication factor, a current, a reaction rate, a dose equivalent rate as well as deposit of energy and recoil energies. For each interesting physical quantity, a Monte-Carlo simulation offers different types of estimators. Tripoli-4 has support for execution in parallel mode. Special features and applications are also presented.

SU-E-T-58: A Novel Monte Carlo Photon Transport Simulation Scheme and Its Application in Cone Beam CT Projection Simulation

International Nuclear Information System (INIS)

Xu, Y; Tian, Z; Jiang, S; Jia, X; Zhou, L

2015-01-01

Purpose: Monte Carlo (MC) simulation is an important tool to solve radiotherapy and medical imaging problems. Low computational efficiency hinders its wide applications. Conventionally, MC is performed in a particle-by -particle fashion. The lack of control on particle trajectory is a main cause of low efficiency in some applications. Take cone beam CT (CBCT) projection simulation as an example, significant amount of computations were wasted on transporting photons that do not reach the detector. To solve this problem, we propose an innovative MC simulation scheme with a path-by-path sampling method. Methods: Consider a photon path starting at the x-ray source. After going through a set of interactions, it ends at the detector. In the proposed scheme, we sampled an entire photon path each time. Metropolis-Hasting algorithm was employed to accept/reject a sampled path based on a calculated acceptance probability, in order to maintain correct relative probabilities among different paths, which are governed by photon transport physics. We developed a package gMMC on GPU with this new scheme implemented. The performance of gMMC was tested in a sample problem of CBCT projection simulation for a homogeneous object. The results were compared to those obtained using gMCDRR, a GPU-based MC tool with the conventional particle-by-particle simulation scheme. Results: Calculated scattered photon signals in gMMC agreed with those from gMCDRR with a relative difference of 3%. It took 3.1 hr. for gMCDRR to simulate 7.8e11 photons and 246.5 sec for gMMC to simulate 1.4e10 paths. Under this setting, both results attained the same ∼2% statistical uncertainty. Hence, a speed-up factor of ∼45.3 was achieved by this new path-by-path simulation scheme, where all the computations were spent on those photons contributing to the detector signal. Conclusion: We innovatively proposed a novel path-by-path simulation scheme that enabled a significant efficiency enhancement for MC particle

SU-E-T-58: A Novel Monte Carlo Photon Transport Simulation Scheme and Its Application in Cone Beam CT Projection Simulation

Energy Technology Data Exchange (ETDEWEB)

Xu, Y [UT Southwestern Medical Center, Dallas, TX (United States); Southern Medical University, Guangzhou (China); Tian, Z; Jiang, S; Jia, X [UT Southwestern Medical Center, Dallas, TX (United States); Zhou, L [Southern Medical University, Guangzhou (China)

2015-06-15

Purpose: Monte Carlo (MC) simulation is an important tool to solve radiotherapy and medical imaging problems. Low computational efficiency hinders its wide applications. Conventionally, MC is performed in a particle-by -particle fashion. The lack of control on particle trajectory is a main cause of low efficiency in some applications. Take cone beam CT (CBCT) projection simulation as an example, significant amount of computations were wasted on transporting photons that do not reach the detector. To solve this problem, we propose an innovative MC simulation scheme with a path-by-path sampling method. Methods: Consider a photon path starting at the x-ray source. After going through a set of interactions, it ends at the detector. In the proposed scheme, we sampled an entire photon path each time. Metropolis-Hasting algorithm was employed to accept/reject a sampled path based on a calculated acceptance probability, in order to maintain correct relative probabilities among different paths, which are governed by photon transport physics. We developed a package gMMC on GPU with this new scheme implemented. The performance of gMMC was tested in a sample problem of CBCT projection simulation for a homogeneous object. The results were compared to those obtained using gMCDRR, a GPU-based MC tool with the conventional particle-by-particle simulation scheme. Results: Calculated scattered photon signals in gMMC agreed with those from gMCDRR with a relative difference of 3%. It took 3.1 hr. for gMCDRR to simulate 7.8e11 photons and 246.5 sec for gMMC to simulate 1.4e10 paths. Under this setting, both results attained the same ∼2% statistical uncertainty. Hence, a speed-up factor of ∼45.3 was achieved by this new path-by-path simulation scheme, where all the computations were spent on those photons contributing to the detector signal. Conclusion: We innovatively proposed a novel path-by-path simulation scheme that enabled a significant efficiency enhancement for MC particle

MVP/GMVP version 3. General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

International Nuclear Information System (INIS)

Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

2017-03-01

In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William

2005-09-01

ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 5.0, the latest version of ITS, contains (1) improvements to the ITS 3.0 continuous-energy codes, (2) multigroup codes with adjoint transport capabilities, (3) parallel implementations of all ITS codes, (4) a general purpose geometry engine for linking with CAD or other geometry formats, and (5) the Cholla facet geometry library. Moreover, the general user friendliness of the software has been enhanced through increased internal error checking and improved code portability.

Simulation of photon transport in a realistic human body model

International Nuclear Information System (INIS)

Baccarne, V.; Turzo, A.; Bizais, Y.; Farine, M.

1997-01-01

A Monte-Carlo photon transport code to simulate scintigraphy is developed. The scintigraphy consists of injecting a patient with a radioactive tracer (Tc, a 140 keV photon emitter) attached to a biologically active molecule. Complicated physical phenomena, photon interactions, occurring in between the radioactive source emission and the detection of the photon on the gamma-camera, require an accurate description. All these phenomena are very sensitive to the characteristics of human tissues and we had to use segmented computerized tomography slices. A preliminary theoretical study of the physical characteristics (rather badly known) of the biological tissues resulted in a two family classification: soft and bone tissues. By devising a Monte-Carlo simulator a systematic investigation was carried out concerning the relative weight of different types of interaction taking place in the traversed tissue. The importance of bone tissues was evidenced in comparison with the soft tissues, as well as the instability of these phenomena as a function of the patient morphology. These information are crucial in the elaboration and validation of correction techniques applied to the diagnosis images of clinical examinations

SU-E-T-112: An OpenCL-Based Cross-Platform Monte Carlo Dose Engine (oclMC) for Coupled Photon-Electron Transport

International Nuclear Information System (INIS)

Tian, Z; Shi, F; Folkerts, M; Qin, N; Jiang, S; Jia, X

2015-01-01

Purpose: Low computational efficiency of Monte Carlo (MC) dose calculation impedes its clinical applications. Although a number of MC dose packages have been developed over the past few years, enabling fast MC dose calculations, most of these packages were developed under NVidia’s CUDA environment. This limited their code portability to other platforms, hindering the introduction of GPU-based MC dose engines to clinical practice. To solve this problem, we developed a cross-platform fast MC dose engine named oclMC under OpenCL environment for external photon and electron radiotherapy. Methods: Coupled photon-electron simulation was implemented with standard analogue simulation scheme for photon transport and Class II condensed history scheme for electron transport. We tested the accuracy and efficiency of oclMC by comparing the doses calculated using oclMC and gDPM, a previously developed GPU-based MC code on NVidia GPU platform, for a 15MeV electron beam and a 6MV photon beam in a homogenous water phantom, a water-bone-lung-water slab phantom and a half-slab phantom. We also tested code portability of oclMC on different devices, including an NVidia GPU, two AMD GPUs and an Intel CPU. Results: Satisfactory agreements were observed in all photon and electron cases, with ∼0.48%–0.53% average dose differences at regions within 10% isodose line for electron beam cases and ∼0.15%–0.17% for photon beam cases. It took oclMC 3–4 sec to perform transport simulation for electron beam on NVidia Titan GPU and 35–51 sec for photon beam, both with ∼0.5% statistical uncertainty. The computation was 6%–17% slower than gDPM due to the differences in both physics model and development environment, which is considered not significant for clinical applications. In terms of code portability, gDPM only runs on NVidia GPUs, while oclMC successfully runs on all the tested devices. Conclusion: oclMC is an accurate and fast MC dose engine. Its high cross

SU-E-T-112: An OpenCL-Based Cross-Platform Monte Carlo Dose Engine (oclMC) for Coupled Photon-Electron Transport

Energy Technology Data Exchange (ETDEWEB)

Tian, Z; Shi, F; Folkerts, M; Qin, N; Jiang, S; Jia, X [The University of Texas Southwestern Medical Ctr, Dallas, TX (United States)

2015-06-15

Purpose: Low computational efficiency of Monte Carlo (MC) dose calculation impedes its clinical applications. Although a number of MC dose packages have been developed over the past few years, enabling fast MC dose calculations, most of these packages were developed under NVidia’s CUDA environment. This limited their code portability to other platforms, hindering the introduction of GPU-based MC dose engines to clinical practice. To solve this problem, we developed a cross-platform fast MC dose engine named oclMC under OpenCL environment for external photon and electron radiotherapy. Methods: Coupled photon-electron simulation was implemented with standard analogue simulation scheme for photon transport and Class II condensed history scheme for electron transport. We tested the accuracy and efficiency of oclMC by comparing the doses calculated using oclMC and gDPM, a previously developed GPU-based MC code on NVidia GPU platform, for a 15MeV electron beam and a 6MV photon beam in a homogenous water phantom, a water-bone-lung-water slab phantom and a half-slab phantom. We also tested code portability of oclMC on different devices, including an NVidia GPU, two AMD GPUs and an Intel CPU. Results: Satisfactory agreements were observed in all photon and electron cases, with ∼0.48%–0.53% average dose differences at regions within 10% isodose line for electron beam cases and ∼0.15%–0.17% for photon beam cases. It took oclMC 3–4 sec to perform transport simulation for electron beam on NVidia Titan GPU and 35–51 sec for photon beam, both with ∼0.5% statistical uncertainty. The computation was 6%–17% slower than gDPM due to the differences in both physics model and development environment, which is considered not significant for clinical applications. In terms of code portability, gDPM only runs on NVidia GPUs, while oclMC successfully runs on all the tested devices. Conclusion: oclMC is an accurate and fast MC dose engine. Its high cross

Two improved Monte Carlo photon cross section techniques

International Nuclear Information System (INIS)

Scudiere, M.B.

1978-01-01

Truncated series of Legendre coefficients and polynomials are often used in multigroup transport computer codes to describe group-to-group angular density transfer functions. Imposition of group structure on the energy continuum may create discontinuities in the first derivative of these functions. Because of the nature of these discontinuities efficient and accurate full-range polynomial expansions are not practically obtainable. Two separate and distinct methods for Monte Carlo photon transport are presented which eliminate essentially all major disadvantages of truncated expansions. In the first method, partial-range expansions are applied between the discontinuities. Here accurate low-order representations are obtained, which yield modest savings in computer charges. The second method employs unique properties of functions to replace them with a few smooth well-behaved representations. This method brings about a considerable savings in computer memory requirements. In addition, accuracy of the first method is maintained, while execution times are reduced even further

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

Science.gov (United States)

García Muñoz, A.; Mills, F. P.

2017-08-01

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

Modelling of electron contamination in clinical photon beams for Monte Carlo dose calculation

International Nuclear Information System (INIS)

Yang, J; Li, J S; Qin, L; Xiong, W; Ma, C-M

2004-01-01

The purpose of this work is to model electron contamination in clinical photon beams and to commission the source model using measured data for Monte Carlo treatment planning. In this work, a planar source is used to represent the contaminant electrons at a plane above the upper jaws. The source size depends on the dimensions of the field size at the isocentre. The energy spectra of the contaminant electrons are predetermined using Monte Carlo simulations for photon beams from different clinical accelerators. A 'random creep' method is employed to derive the weight of the electron contamination source by matching Monte Carlo calculated monoenergetic photon and electron percent depth-dose (PDD) curves with measured PDD curves. We have integrated this electron contamination source into a previously developed multiple source model and validated the model for photon beams from Siemens PRIMUS accelerators. The EGS4 based Monte Carlo user code BEAM and MCSIM were used for linac head simulation and dose calculation. The Monte Carlo calculated dose distributions were compared with measured data. Our results showed good agreement (less than 2% or 2 mm) for 6, 10 and 18 MV photon beams

MVP/GMVP Version 3. General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods (Translated document)

International Nuclear Information System (INIS)

Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

2017-03-01

In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)

Generalized free-space diffuse photon transport model based on the influence analysis of a camera lens diaphragm.

Science.gov (United States)

Chen, Xueli; Gao, Xinbo; Qu, Xiaochao; Chen, Duofang; Ma, Xiaopeng; Liang, Jimin; Tian, Jie

2010-10-10

The camera lens diaphragm is an important component in a noncontact optical imaging system and has a crucial influence on the images registered on the CCD camera. However, this influence has not been taken into account in the existing free-space photon transport models. To model the photon transport process more accurately, a generalized free-space photon transport model is proposed. It combines Lambertian source theory with analysis of the influence of the camera lens diaphragm to simulate photon transport process in free space. In addition, the radiance theorem is also adopted to establish the energy relationship between the virtual detector and the CCD camera. The accuracy and feasibility of the proposed model is validated with a Monte-Carlo-based free-space photon transport model and physical phantom experiment. A comparison study with our previous hybrid radiosity-radiance theorem based model demonstrates the improvement performance and potential of the proposed model for simulating photon transport process in free space.

Lecture 1. Monte Carlo basics. Lecture 2. Adjoint Monte Carlo. Lecture 3. Coupled Forward-Adjoint calculations

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J.E. [Delft University of Technology, Interfaculty Reactor Institute, Delft (Netherlands)

2000-07-01

The Monte Carlo method is a statistical method to solve mathematical and physical problems using random numbers. The principle of the methods will be demonstrated for a simple mathematical problem and for neutron transport. Various types of estimators will be discussed, as well as generally applied variance reduction methods like splitting, Russian roulette and importance biasing. The theoretical formulation for solving eigenvalue problems for multiplying systems will be shown. Some reflections will be given about the applicability of the Monte Carlo method, its limitations and its future prospects for reactor physics calculations. Adjoint Monte Carlo is a Monte Carlo game to solve the adjoint neutron (or photon) transport equation. The adjoint transport equation can be interpreted in terms of simulating histories of artificial particles, which show properties of neutrons that move backwards in history. These particles will start their history at the detector from which the response must be estimated and give a contribution to the estimated quantity when they hit or pass through the neutron source. Application to multigroup transport formulation will be demonstrated Possible implementation for the continuous energy case will be outlined. The inherent advantages and disadvantages of the method will be discussed. The Midway Monte Carlo method will be presented for calculating a detector response due to a (neutron or photon) source. A derivation will be given of the basic formula for the Midway Monte Carlo method The black absorber technique, allowing for a cutoff of particle histories when reaching the midway surface in one of the calculations will be derived. An extension of the theory to coupled neutron-photon problems is given. The method will be demonstrated for an oil well logging problem, comprising a neutron source in a borehole and photon detectors to register the photons generated by inelastic neutron scattering. (author)

Lecture 1. Monte Carlo basics. Lecture 2. Adjoint Monte Carlo. Lecture 3. Coupled Forward-Adjoint calculations

International Nuclear Information System (INIS)

Hoogenboom, J.E.

2000-01-01

The Monte Carlo method is a statistical method to solve mathematical and physical problems using random numbers. The principle of the methods will be demonstrated for a simple mathematical problem and for neutron transport. Various types of estimators will be discussed, as well as generally applied variance reduction methods like splitting, Russian roulette and importance biasing. The theoretical formulation for solving eigenvalue problems for multiplying systems will be shown. Some reflections will be given about the applicability of the Monte Carlo method, its limitations and its future prospects for reactor physics calculations. Adjoint Monte Carlo is a Monte Carlo game to solve the adjoint neutron (or photon) transport equation. The adjoint transport equation can be interpreted in terms of simulating histories of artificial particles, which show properties of neutrons that move backwards in history. These particles will start their history at the detector from which the response must be estimated and give a contribution to the estimated quantity when they hit or pass through the neutron source. Application to multigroup transport formulation will be demonstrated Possible implementation for the continuous energy case will be outlined. The inherent advantages and disadvantages of the method will be discussed. The Midway Monte Carlo method will be presented for calculating a detector response due to a (neutron or photon) source. A derivation will be given of the basic formula for the Midway Monte Carlo method The black absorber technique, allowing for a cutoff of particle histories when reaching the midway surface in one of the calculations will be derived. An extension of the theory to coupled neutron-photon problems is given. The method will be demonstrated for an oil well logging problem, comprising a neutron source in a borehole and photon detectors to register the photons generated by inelastic neutron scattering. (author)

ScintSim1: a new Monte Carlo simulation code for transport of optical photons in 2D arrays of scintillation detectors

International Nuclear Information System (INIS)

Mosleh-Shirazi, Mohammad Amin; Karbasi, Sareh; Zarrini-Monfared, Zinat; Zamani, Ali

2014-01-01

Two-dimensional (2D) arrays of thick segmented scintillators are of interest as X-ray detectors for both 2D and 3D image-guided radiotherapy (IGRT). Their detection process involves ionizing radiation energy deposition followed by production and transport of optical photons. Only a very limited number of optical Monte Carlo simulation models exist, which has limited the number of modeling studies that have considered both stages of the detection process. We present ScintSim1, an in-house optical Monte Carlo simulation code for 2D arrays of scintillation crystals, developed in the MATLAB programming environment. The code was rewritten and revised based on an existing program for single-element detectors, with the additional capability to model 2D arrays of elements with configurable dimensions, material, etc., The code generates and follows each optical photon history through the detector element (and, in case of cross-talk, the surrounding ones) until it reaches a configurable receptor, or is attenuated. The new model was verified by testing against relevant theoretically known behaviors or quantities and the results of a validated single-element model. For both sets of comparisons, the discrepancies in the calculated quantities were all <1%. The results validate the accuracy of the new code, which is a useful tool in scintillation detector optimization. (author)

Coupling photon Monte Carlo simulation and CAD software. Application to X-ray nondestructive evaluation

International Nuclear Information System (INIS)

Tabary, J.; Gliere, A.

2001-01-01

A Monte Carlo radiation transport simulation program, EGS Nova, and a computer aided design software, BRL-CAD, have been coupled within the framework of Sindbad, a nondestructive evaluation (NDE) simulation system. In its current status, the program is very valuable in a NDE laboratory context, as it helps simulate the images due to the uncollided and scattered photon fluxes in a single NDE software environment, without having to switch to a Monte Carlo code parameters set. Numerical validations show a good agreement with EGS4 computed and published data. As the program's major drawback is the execution time, computational efficiency improvements are foreseen. (orig.)

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...

Application of an efficient materials perturbation technique to Monte Carlo photon transport calculations in borehole logging

International Nuclear Information System (INIS)

Picton, D.J.; Harris, R.G.; Randle, K.; Weaver, D.R.

1995-01-01

This paper describes a simple, accurate and efficient technique for the calculation of materials perturbation effects in Monte Carlo photon transport calculations. It is particularly suited to the application for which it was developed, namely the modelling of a dual detector density tool as used in borehole logging. However, the method would be appropriate to any photon transport calculation in the energy range 0.1 to 2 MeV, in which the predominant processes are Compton scattering and photoelectric absorption. The method enables a single set of particle histories to provide results for an array of configurations in which material densities or compositions vary. It can calculate the effects of small perturbations very accurately, but is by no means restricted to such cases. For the borehole logging application described here the method has been found to be efficient for a moderate range of variation in the bulk density (of the order of ±30% from a reference value) or even larger changes to a limited portion of the system (e.g. a low density mudcake of the order of a few tens of mm in thickness). The effective speed enhancement over an equivalent set of individual calculations is in the region of an order of magnitude or more. Examples of calculations on a dual detector density tool are given. It is demonstrated that the method predicts, to a high degree of accuracy, the variation of detector count rates with formation density, and that good results are also obtained for the effects of mudcake layers. An interesting feature of the results is that relative count rates (the ratios of count rates obtained with different configurations) can usually be determined more accurately than the absolute values of the count rates. (orig.)

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

Efficient photon treatment planning by the use of Swiss Monte Carlo Plan

International Nuclear Information System (INIS)

Fix, M K; Manser, P; Frei, D; Volken, W; Mini, R; Born, E J

2007-01-01

Currently photon Monte Carlo treatment planning (MCTP) for a patient stored in the patient database of a treatment planning system (TPS) usually can only be performed using a cumbersome multi-step procedure where many user interactions are needed. Automation is needed for usage in clinical routine. In addition, because of the long computing time in MCTP, optimization of the MC calculations is essential. For these purposes a new GUI-based photon MC environment has been developed resulting in a very flexible framework, namely the Swiss Monte Carlo Plan (SMCP). Appropriate MC transport methods are assigned to different geometric regions by still benefiting from the features included in the TPS. In order to provide a flexible MC environment the MC particle transport has been divided into different parts: source, beam modifiers, and patient. The source part includes: Phase space-source, source models, and full MC transport through the treatment head. The beam modifier part consists of one module for each beam modifier. To simulate the radiation transport through each individual beam modifier, one out of three full MC transport codes can be selected independently. Additionally, for each beam modifier a simple or an exact geometry can be chosen. Thereby, different complexity levels of radiation transport are applied during the simulation. For the patient dose calculation two different MC codes are available. A special plug-in in Eclipse providing all necessary information by means of Dicom streams was used to start the developed MC GUI. The implementation of this framework separates the MC transport from the geometry and the modules pass the particles in memory, hence no files are used as interface. The implementation is realized for 6 and 15 MV beams of a Varian Clinac 2300 C/D. Several applications demonstrate the usefulness of the framework. Apart from applications dealing with the beam modifiers, three patient cases are shown. Thereby, comparisons between MC

Vectorization of Monte Carlo particle transport

International Nuclear Information System (INIS)

Burns, P.J.; Christon, M.; Schweitzer, R.; Lubeck, O.M.; Wasserman, H.J.; Simmons, M.L.; Pryor, D.V.

1989-01-01

This paper reports that fully vectorized versions of the Los Alamos National Laboratory benchmark code Gamteb, a Monte Carlo photon transport algorithm, were developed for the Cyber 205/ETA-10 and Cray X-MP/Y-MP architectures. Single-processor performance measurements of the vector and scalar implementations were modeled in a modified Amdahl's Law that accounts for additional data motion in the vector code. The performance and implementation strategy of the vector codes are related to architectural features of each machine. Speedups between fifteen and eighteen for Cyber 205/ETA-10 architectures, and about nine for CRAY X-MP/Y-MP architectures are observed. The best single processor execution time for the problem was 0.33 seconds on the ETA-10G, and 0.42 seconds on the CRAY Y-MP

Temperature variance study in Monte-Carlo photon transport theory

International Nuclear Information System (INIS)

Giorla, J.

1985-10-01

We study different Monte-Carlo methods for solving radiative transfer problems, and particularly Fleck's Monte-Carlo method. We first give the different time-discretization schemes and the corresponding stability criteria. Then we write the temperature variance as a function of the variances of temperature and absorbed energy at the previous time step. Finally we obtain some stability criteria for the Monte-Carlo method in the stationary case [fr

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

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...

SANDYL, 3-D Time-Dependent and Space-Dependent Gamma Electron Cascade Transport by Monte-Carlo

International Nuclear Information System (INIS)

Haggmark, L.G.

1980-01-01

1 - Description of problem or function: SANDYL performs three- dimensional, time and space dependent Monte Carlo transport calculations for photon-electron cascades in complex systems. 2 - Method of solution: The problem geometry is divided into zones of homogeneous atomic composition bounded by sections of planes and quadrics. The material of each zone is a specified element or combination of elements. For a photon history, the trajectory is generated by following the photon from scattering to scattering using the various probability distributions to find distances between collisions, types of collisions, types of secondaries, and their energies and scattering angles. The photon interactions are photoelectric absorption (atomic ionization), coherent scattering, incoherent scattering, and pair production. The secondary photons which are followed include Bremsstrahlung, fluorescence photons, and positron-electron annihilation radiation. The condensed-history Monte Carlo method is used for the electron transport. In a history, the spatial steps taken by an electron are pre-computed and may include the effects of a number of collisions. The corresponding scattering angle and energy loss in the step are found from the multiple scattering distributions of these quantities. Atomic ionization and secondary particles are generated with the step according to the probabilities for their occurrence. Electron energy loss is through inelastic electron-electron collisions, Bremsstrahlung generation, and polarization of the medium (density effect). Included in the loss is the fluctuation due to the variation in the number of energy-loss collisions in a given Monte Carlo step (straggling). Scattering angular distributions are determined from elastic nuclear-collision cross sections corrected for electron-electron interactions. The secondary electrons which are followed included knock-on, pair, Auger (through atomic ionizations), Compton, and photoelectric electrons. 3

An efficient framework for photon Monte Carlo treatment planning

International Nuclear Information System (INIS)

Fix, Michael K; Manser, Peter; Frei, Daniel; Volken, Werner; Mini, Roberto; Born, Ernst J

2007-01-01

Currently photon Monte Carlo treatment planning (MCTP) for a patient stored in the patient database of a treatment planning system (TPS) can usually only be performed using a cumbersome multi-step procedure where many user interactions are needed. This means automation is needed for usage in clinical routine. In addition, because of the long computing time in MCTP, optimization of the MC calculations is essential. For these purposes a new graphical user interface (GUI)-based photon MC environment has been developed resulting in a very flexible framework. By this means appropriate MC transport methods are assigned to different geometric regions by still benefiting from the features included in the TPS. In order to provide a flexible MC environment, the MC particle transport has been divided into different parts: the source, beam modifiers and the patient. The source part includes the phase-space source, source models and full MC transport through the treatment head. The beam modifier part consists of one module for each beam modifier. To simulate the radiation transport through each individual beam modifier, one out of three full MC transport codes can be selected independently. Additionally, for each beam modifier a simple or an exact geometry can be chosen. Thereby, different complexity levels of radiation transport are applied during the simulation. For the patient dose calculation, two different MC codes are available. A special plug-in in Eclipse providing all necessary information by means of Dicom streams was used to start the developed MC GUI. The implementation of this framework separates the MC transport from the geometry and the modules pass the particles in memory; hence, no files are used as the interface. The implementation is realized for 6 and 15 MV beams of a Varian Clinac 2300 C/D. Several applications demonstrate the usefulness of the framework. Apart from applications dealing with the beam modifiers, two patient cases are shown. Thereby

Monte Carlo simulation of two-photon processes

International Nuclear Information System (INIS)

Daverveldt, P.H.W.M.

1985-01-01

During the last two decades e + e - collider experiments provided physicists with a wealth of important discoveries concerning elementary particle physics. This thesis explains in detail how the Monte Carlo approach can be applied to establish the comparison between two-photon experiments and theory. The author describes the main motives for and objectives of two-photon research. He defines the kinematics and pays attention to some special kinematical regions. Also a popular approximation for the exact differential cross section is reviewed. Next he discusses the calculation of the complete lowest order cross section for processes with four leptons in the final state and for reactions such as e + e - →e + e - qanti q, e + e - →μ + μ - qanti q. Radiative corrections to the multiperipheral diagrams are considered. The author explains in detail the distinction between soft and hard photon corrections which turns out to be somewhat more tricky than in the case of radiative corrections to one-photon processes. Finally, he presents some results which were obtained by using the event generators. (Auth.)

Study on Photon Transport Problem Based on the Platform of Molecular Optical Simulation Environment

Directory of Open Access Journals (Sweden)

Kuan Peng

2010-01-01

Full Text Available As an important molecular imaging modality, optical imaging has attracted increasing attention in the recent years. Since the physical experiment is usually complicated and expensive, research methods based on simulation platforms have obtained extensive attention. We developed a simulation platform named Molecular Optical Simulation Environment (MOSE to simulate photon transport in both biological tissues and free space for optical imaging based on noncontact measurement. In this platform, Monte Carlo (MC method and the hybrid radiosity-radiance theorem are used to simulate photon transport in biological tissues and free space, respectively, so both contact and noncontact measurement modes of optical imaging can be simulated properly. In addition, a parallelization strategy for MC method is employed to improve the computational efficiency. In this paper, we study the photon transport problems in both biological tissues and free space using MOSE. The results are compared with Tracepro, simplified spherical harmonics method (SPn, and physical measurement to verify the performance of our study method on both accuracy and efficiency.

Study on photon transport problem based on the platform of molecular optical simulation environment.

Science.gov (United States)

Peng, Kuan; Gao, Xinbo; Liang, Jimin; Qu, Xiaochao; Ren, Nunu; Chen, Xueli; Ma, Bin; Tian, Jie

2010-01-01

As an important molecular imaging modality, optical imaging has attracted increasing attention in the recent years. Since the physical experiment is usually complicated and expensive, research methods based on simulation platforms have obtained extensive attention. We developed a simulation platform named Molecular Optical Simulation Environment (MOSE) to simulate photon transport in both biological tissues and free space for optical imaging based on noncontact measurement. In this platform, Monte Carlo (MC) method and the hybrid radiosity-radiance theorem are used to simulate photon transport in biological tissues and free space, respectively, so both contact and noncontact measurement modes of optical imaging can be simulated properly. In addition, a parallelization strategy for MC method is employed to improve the computational efficiency. In this paper, we study the photon transport problems in both biological tissues and free space using MOSE. The results are compared with Tracepro, simplified spherical harmonics method (SP(n)), and physical measurement to verify the performance of our study method on both accuracy and efficiency.

EGS4, Electron Photon Shower Simulation by Monte-Carlo

International Nuclear Information System (INIS)

1998-01-01

1 - Description of program or function: The EGS code system is one of a chain of three codes designed to solve the electromagnetic shower problem by Monte Carlo simulation. This chain makes possible simulation of almost any electron-photon transport problem conceivable. The structure of the system, with its global features, modular form, and structured programming, is readily adaptable to virtually any interfacing scheme that is desired on the part of the user. EGS4 is a package of subroutines plus block data with a flexible user interface. This allows for greater flexibility without requiring the user to be overly familiar with the internal details of the code. Combining this with the macro facility capabilities of the Mortran3 language, this reduces the likelihood that user edits will introduce bugs into the code. EGS4 uses material cross section and branching ratio data created and fit by the companion code, PEGS4. EGS4 allows for the implementation of importance sampling and other variance reduction techniques such as leading particle biasing, splitting, path length biasing, Russian roulette, etc. 2 - Method of solution: EGS employs the Monte Carlo method of solution. It allows all of the fundamental processes to be included and arbitrary geometries can be treated, also. Other minor processes, such as photoneutron production, can be added as a further generalization. Since showers develop randomly according to the quantum laws of probability, each shower is different. We again are led to the Monte Carlo method. 3 - Restrictions on the complexity of the problem: None noted

MCNP: Photon benchmark problems

International Nuclear Information System (INIS)

Whalen, D.J.; Hollowell, D.E.; Hendricks, J.S.

1991-09-01

The recent widespread, markedly increased use of radiation transport codes has produced greater user and institutional demand for assurance that such codes give correct results. Responding to these pressing requirements for code validation, the general purpose Monte Carlo transport code MCNP has been tested on six different photon problem families. MCNP was used to simulate these six sets numerically. Results for each were compared to the set's analytical or experimental data. MCNP successfully predicted the analytical or experimental results of all six families within the statistical uncertainty inherent in the Monte Carlo method. From this we conclude that MCNP can accurately model a broad spectrum of photon transport problems. 8 refs., 30 figs., 5 tabs

Application of discrete ordinates and Monte Carlo methods to transport of photons from environmental sources

International Nuclear Information System (INIS)

Ryman, J.C.; Eckerman, K.F.; Shultis, J.K.; Faw, R.E.; Dillman, L.T.

1996-01-01

Federal Guidance Report No. 12 tabulates dose coefficients for external exposure to photons and electrons emitted by radionuclides distributed in air, water, and soil. Although the dose coefficients of this report are based on previously developed dosimetric methodologies, they are derived from new, detailed calculations of energy and angular distributions of the radiations incident on the body and the transport of these radiations within the body. Effort was devoted to expanding the information available for assessment of radiation dose from radionuclides distributed on or below the surface of the ground. A companion paper (External Exposure to Radionuclides in Air, Water, and Soil) discusses the significance of the new tabulations of coefficients and provides detiled comparisons to previously published values. This paper discusses details of the photon transport calculations

DPM, a fast, accurate Monte Carlo code optimized for photon and electron radiotherapy treatment planning dose calculations

International Nuclear Information System (INIS)

Sempau, Josep; Wilderman, Scott J.; Bielajew, Alex F.

2000-01-01

A new Monte Carlo (MC) algorithm, the 'dose planning method' (DPM), and its associated computer program for simulating the transport of electrons and photons in radiotherapy class problems employing primary electron beams, is presented. DPM is intended to be a high-accuracy MC alternative to the current generation of treatment planning codes which rely on analytical algorithms based on an approximate solution of the photon/electron Boltzmann transport equation. For primary electron beams, DPM is capable of computing 3D dose distributions (in 1 mm 3 voxels) which agree to within 1% in dose maximum with widely used and exhaustively benchmarked general-purpose public-domain MC codes in only a fraction of the CPU time. A representative problem, the simulation of 1 million 10 MeV electrons impinging upon a water phantom of 128 3 voxels of 1 mm on a side, can be performed by DPM in roughly 3 min on a modern desktop workstation. DPM achieves this performance by employing transport mechanics and electron multiple scattering distribution functions which have been derived to permit long transport steps (of the order of 5 mm) which can cross heterogeneity boundaries. The underlying algorithm is a 'mixed' class simulation scheme, with differential cross sections for hard inelastic collisions and bremsstrahlung events described in an approximate manner to simplify their sampling. The continuous energy loss approximation is employed for energy losses below some predefined thresholds, and photon transport (including Compton, photoelectric absorption and pair production) is simulated in an analogue manner. The δ-scattering method (Woodcock tracking) is adopted to minimize the computational costs of transporting photons across voxels. (author)

Parallel processing of Monte Carlo code MCNP for particle transport problem

Energy Technology Data Exchange (ETDEWEB)

Higuchi, Kenji; Kawasaki, Takuji

1996-06-01

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

Parallelization of MCNP 4, a Monte Carlo neutron and photon transport code system, in highly parallel distributed memory type computer

International Nuclear Information System (INIS)

Masukawa, Fumihiro; Takano, Makoto; Naito, Yoshitaka; Yamazaki, Takao; Fujisaki, Masahide; Suzuki, Koichiro; Okuda, Motoi.

1993-11-01

In order to improve the accuracy and calculating speed of shielding analyses, MCNP 4, a Monte Carlo neutron and photon transport code system, has been parallelized and measured of its efficiency in the highly parallel distributed memory type computer, AP1000. The code has been analyzed statically and dynamically, then the suitable algorithm for parallelization has been determined for the shielding analysis functions of MCNP 4. This includes a strategy where a new history is assigned to the idling processor element dynamically during the execution. Furthermore, to avoid the congestion of communicative processing, the batch concept, processing multi-histories by a unit, has been introduced. By analyzing a sample cask problem with 2,000,000 histories by the AP1000 with 512 processor elements, the 82 % of parallelization efficiency is achieved, and the calculational speed has been estimated to be around 50 times as fast as that of FACOM M-780. (author)

Monte Carlo simulation of gas Cerenkov detectors

International Nuclear Information System (INIS)

Mack, J.M.; Jain, M.; Jordan, T.M.

1984-01-01

Theoretical study of selected gamma-ray and electron diagnostic necessitates coupling Cerenkov radiation to electron/photon cascades. A Cerenkov production model and its incorporation into a general geometry Monte Carlo coupled electron/photon transport code is discussed. A special optical photon ray-trace is implemented using bulk optical properties assigned to each Monte Carlo zone. Good agreement exists between experimental and calculated Cerenkov data in the case of a carbon-dioxide gas Cerenkov detector experiment. Cerenkov production and threshold data are presented for a typical carbon-dioxide gas detector that converts a 16.7 MeV photon source to Cerenkov light, which is collected by optics and detected by a photomultiplier

Comparison of Monte Carlo simulations of photon/electron dosimetry in microscale applications

International Nuclear Information System (INIS)

Poneja, O.P.; Chawla, R.; Negreanu, C.; Stepanek, J.

2003-01-01

It is important to establish reliable calculational tools to plan and analyse representative microdosimetry experiments in the context of microbeam radiation therapy development. In this paper, an attempt has been made to investigate the suitability of the MCNP4C Monte Carlo code to adequately model photon/electron transport over micron distances. The case of a single cylindrical microbeam of 25-micron diameter incident on a water phantom has been simulated in detail with both MCNP4C and the code PSI-GEANT, for different incident photon energies, to get absorbed dose distributions at various depths, with and without electron transport being considered. In addition, dose distributions calculated for a single microbeam with a photon spectrum representative of the European Synchrotron Radiation Facility (ESRF) have been compared. Finally, a large number of cylindrical microbeams ( a total of 2601 beams, placed on a 200-micron square pitch, covering an area of lcm 2 ) incident on a water phantom have been considered to study cumulative radial dose distributions at different depths. From these distributions, ratios of peak (within the microbeam) to valley (mid-point along the diagonal connecting two microbeams) dose values have been determined. The various comparisons with PSI-GEANT results have shown that MCNP4C, with its high flexibility in terms of its numerous source and geometry description options, variance reduction methods, detailed error analysis, statistical checks and different tally types, can be a valuable tool for the analysis of microbeam experiments. Copyright (2003) Australasian College of Physical Scientists and Engineers in Medicine

The Monte Carlo simulation of the Ladon photon beam facility

International Nuclear Information System (INIS)

Strangio, C.

1976-01-01

The backward compton scattering of laser light against high energy electrons has been simulated with a Monte Carlo method. The main features of the produced photon beam are reported as well as a careful description of the numerical calculation

Study on Photon Transport Problem Based on the Platform of Molecular Optical Simulation Environment

Science.gov (United States)

Peng, Kuan; Gao, Xinbo; Liang, Jimin; Qu, Xiaochao; Ren, Nunu; Chen, Xueli; Ma, Bin; Tian, Jie

2010-01-01

As an important molecular imaging modality, optical imaging has attracted increasing attention in the recent years. Since the physical experiment is usually complicated and expensive, research methods based on simulation platforms have obtained extensive attention. We developed a simulation platform named Molecular Optical Simulation Environment (MOSE) to simulate photon transport in both biological tissues and free space for optical imaging based on noncontact measurement. In this platform, Monte Carlo (MC) method and the hybrid radiosity-radiance theorem are used to simulate photon transport in biological tissues and free space, respectively, so both contact and noncontact measurement modes of optical imaging can be simulated properly. In addition, a parallelization strategy for MC method is employed to improve the computational efficiency. In this paper, we study the photon transport problems in both biological tissues and free space using MOSE. The results are compared with Tracepro, simplified spherical harmonics method (S P n), and physical measurement to verify the performance of our study method on both accuracy and efficiency. PMID:20445737

MVP/GMVP 2: general purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

International Nuclear Information System (INIS)

Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki

2005-06-01

In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector super-computers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, (5) continuous-energy calculation at arbitrary temperatures, (6) estimation of real variances in eigenvalue problems, (7) point detector and surface crossing estimators, (8) statistical geometry model, (9) function of reactor noise analysis (simulation of the Feynman-α experiment), (10) arbitrary shaped lattice boundary, (11) periodic boundary condition, (12) parallelization with standard libraries (MPI, PVM), (13) supporting many platforms, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them. (author)

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

International Nuclear Information System (INIS)

Martin, William R.; Brown, Forrest B.

2001-01-01

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

Neutron-induced photon production in MCNP

International Nuclear Information System (INIS)

Little, R.C.; Seamon, R.E.

1983-01-01

An improved method of neutron-induced photon production has been incorporated into the Monte Carlo transport code MCNP. The new method makes use of all partial photon-production reaction data provided by ENDF/B evaluators including photon-production cross sections as well as energy and angular distributions of secondary photons. This faithful utilization of sophisticated ENDF/B evaluations allows more precise MCNP calculations for several classes of coupled neutron-photon problems

Exponential convergence on a continuous Monte Carlo transport problem

International Nuclear Information System (INIS)

Booth, T.E.

1997-01-01

For more than a decade, it has been known that exponential convergence on discrete transport problems was possible using adaptive Monte Carlo techniques. An adaptive Monte Carlo method that empirically produces exponential convergence on a simple continuous transport problem is described

Fiber transport of spatially entangled photons

Science.gov (United States)

Löffler, W.; Eliel, E. R.; Woerdman, J. P.; Euser, T. G.; Scharrer, M.; Russell, P.

2012-03-01

High-dimensional entangled photons pairs are interesting for quantum information and cryptography: Compared to the well-known 2D polarization case, the stronger non-local quantum correlations could improve noise resistance or security, and the larger amount of information per photon increases the available bandwidth. One implementation is to use entanglement in the spatial degree of freedom of twin photons created by spontaneous parametric down-conversion, which is equivalent to orbital angular momentum entanglement, this has been proven to be an excellent model system. The use of optical fiber technology for distribution of such photons has only very recently been practically demonstrated and is of fundamental and applied interest. It poses a big challenge compared to the established time and frequency domain methods: For spatially entangled photons, fiber transport requires the use of multimode fibers, and mode coupling and intermodal dispersion therein must be minimized not to destroy the spatial quantum correlations. We demonstrate that these shortcomings of conventional multimode fibers can be overcome by using a hollow-core photonic crystal fiber, which follows the paradigm to mimic free-space transport as good as possible, and are able to confirm entanglement of the fiber-transported photons. Fiber transport of spatially entangled photons is largely unexplored yet, therefore we discuss the main complications, the interplay of intermodal dispersion and mode mixing, the influence of external stress and core deformations, and consider the pros and cons of various fiber types.

New model for mines and transportation tunnels external dose calculation using Monte Carlo simulation

International Nuclear Information System (INIS)

Allam, Kh. A.

2017-01-01

In this work, a new methodology is developed based on Monte Carlo simulation for tunnels and mines external dose calculation. Tunnels external dose evaluation model of a cylindrical shape of finite thickness with an entrance and with or without exit. A photon transportation model was applied for exposure dose calculations. A new software based on Monte Carlo solution was designed and programmed using Delphi programming language. The variation of external dose due to radioactive nuclei in a mine tunnel and the corresponding experimental data lies in the range 7.3 19.9%. The variation of specific external dose rate with position in, tunnel building material density and composition were studied. The given new model has more flexible for real external dose in any cylindrical tunnel structure calculations. (authors)

TRIPOLI-4.3.3 and 4.4, Coupled Neutron, Photon, Electron, Positron 3-D, Time Dependent Monte-Carlo, Transport Calculation

International Nuclear Information System (INIS)

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

2008-01-01

1 - Description of program or function: TRIPOLI-4 is a general purpose radiation transport code. It uses the Monte Carlo method to simulate neutron and photon behaviour in three-dimensional geometries. The main areas of applications include but are not restricted to: radiation protection and shielding, nuclear criticality safety, fission and fusion reactor design, nuclear instrumentation. In addition, it can simulate electron-photon cascade showers. It computes particle fluxes and currents and several related physical quantities such as, reaction rates, dose rates, heating, energy deposition, effective multiplication factor, perturbation effects due to density, concentration or partial cross-section variations. The summary precises the types of particles, the nuclear data format and cross sections, the energy ranges, the geometry, the sources, the calculated physical quantities and estimators, the biasing, the time-dependant transport for neutrons, the perturbation, the coupled particle transport and the qualification benchmarks. Data libraries distributed with the TRIPOLI-4: ENDFB6R4, ENDL, JEF2, Mott-Rutherford and Qfission. NEA-1716/04: TRIPOLI-4.4 does not contain the source programs. New features available in TRIPOLI-4 version 4 concern the following points: New biasing features, neutron collision in multigroup homogenized mode, display of the collision sites, ENDF format evaluations, computation of the gamma source produced by neutrons, output format for all results, Verbose level for output warnings, photons reactions rates, XML format output, ENDF format evaluations, combinatorial geometry checks, Green's functions files, and neutronics-shielding coupling. 2 - Methods: The geometry package allows the user to describe a three dimensional configuration by means of surfaces (as in the MCNP code) and also through predefined shapes combine with operators (union, intersection, subtraction...). It is also possible to repeat a pattern to built a network of networks

Discrete Diffusion Monte Carlo for Electron Thermal Transport

Science.gov (United States)

Chenhall, Jeffrey; Cao, Duc; Wollaeger, Ryan; Moses, Gregory

2014-10-01

The iSNB (implicit Schurtz Nicolai Busquet electron thermal transport method of Cao et al. is adapted to a Discrete Diffusion Monte Carlo (DDMC) solution method for eventual inclusion in a hybrid IMC-DDMC (Implicit Monte Carlo) method. The hybrid method will combine the efficiency of a diffusion method in short mean free path regions with the accuracy of a transport method in long mean free path regions. The Monte Carlo nature of the approach allows the algorithm to be massively parallelized. Work to date on the iSNB-DDMC method will be presented. This work was supported by Sandia National Laboratory - Albuquerque.

Application of the Monte Carlo method to the study of the response of an organic liquid scintillator irradiated by photons

International Nuclear Information System (INIS)

Dupre, Corinne.

1982-10-01

The Monte Carlo method was applied to simulate the transport of a photon beam in an organic liquid scintillation detector. The interactions of secondary gamma rays and electrons with the detector and its peripheral materials components such as the pyrex glass container are included. The pulse height spectra and the detectors efficiency are compared with calculated and measured results. Calculations and programmation methods are presented in the same way as results concerning cobalt and cesium sources [fr

Analysis of EBR-II neutron and photon physics by multidimensional transport-theory techniques

International Nuclear Information System (INIS)

Jacqmin, R.P.; Finck, P.J.; Palmiotti, G.

1994-01-01

This paper contains a review of the challenges specific to the EBR-II core physics, a description of the methods and techniques which have been developed for addressing these challenges, and the results of some validation studies relative to power-distribution calculations. Numerical tests have shown that the VARIANT nodal code yields eigenvalue and power predictions as accurate as finite difference and discrete ordinates transport codes, at a small fraction of the cost. Comparisons with continuous-energy Monte Carlo results have proven that the errors introduced by the use of the diffusion-theory approximation in the collapsing procedure to obtain broad-group cross sections, kerma factors, and photon-production matrices, have a small impact on the EBR-II neutron/photon power distribution

Simulation of transport equations with Monte Carlo

International Nuclear Information System (INIS)

Matthes, W.

1975-09-01

The main purpose of the report is to explain the relation between the transport equation and the Monte Carlo game used for its solution. The introduction of artificial particles carrying a weight provides one with high flexibility in constructing many different games for the solution of the same equation. This flexibility opens a way to construct a Monte Carlo game for the solution of the adjoint transport equation. Emphasis is laid mostly on giving a clear understanding of what to do and not on the details of how to do a specific game

Monte Carlo code development in Los Alamos

International Nuclear Information System (INIS)

Carter, L.L.; Cashwell, E.D.; Everett, C.J.; Forest, C.A.; Schrandt, R.G.; Taylor, W.M.; Thompson, W.L.; Turner, G.D.

1974-01-01

The present status of Monte Carlo code development at Los Alamos Scientific Laboratory is discussed. A brief summary is given of several of the most important neutron, photon, and electron transport codes. 17 references. (U.S.)

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

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

Simulation of neutron transport process, photons and charged particles within the Monte Carlo method

International Nuclear Information System (INIS)

Androsenko, A.A.; Androsenko, P.A.; Artamonov, S.N.; Bolonkina, G.V.; Lomtev, V.L.; Pupko, S.V.

1991-01-01

Description is given to the program system BRAND designed for the accurate solution of non-stationary transport equation of neutrons, photons and charged particles in the conditions of real three-dimensional geometry. An extensive set of local and non-local estimates provides an opportunity of calculating a great set of linear functionals normally being of interest in the calculation of reactors, radiation protection and experiment simulation. The process of particle interaction with substance is simulated on the basis of individual non-group data on each isotope of the composition. 24 refs

Single photon transport by a moving atom

International Nuclear Information System (INIS)

Afanasiev, A E; Melentiev, P N; Kuzin, A A; Yu Kalatskiy, A; Balykin, V I

2017-01-01

The results of investigation of photon transport through the subwavelength hole in the opaque screen by using single neutral atom are represented. The basis of the proposed and implemented method is the absorption of a photon by a neutral atom immediately before the subwavelength aperture, traveling of the atoms through the hole and emission of a photon on the other side of the screen. Realized method is the alternative approach to existing for photon transport through a subwavelength aperture: 1) self-sustained transmittance of a photon through the aperture according to the Bethe’s model; 2) extra ordinary transmission because of surface-plasmon excitation. (paper)

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

Performance of three-photon PET imaging: Monte Carlo simulations

International Nuclear Information System (INIS)

Kacperski, Krzysztof; Spyrou, Nicholas M

2005-01-01

We have recently introduced the idea of making use of three-photon positron annihilations in positron emission tomography. In this paper, the basic characteristics of the three-gamma imaging in PET are studied by means of Monte Carlo simulations and analytical computations. Two typical configurations of human and small animal scanners are considered. Three-photon imaging requires high-energy resolution detectors. Parameters currently attainable by CdZnTe semiconductor detectors, the technology of choice for the future development of radiation imaging, are assumed. Spatial resolution is calculated as a function of detector energy resolution and size, position in the field of view, scanner size and the energies of the three-gamma annihilation photons. Possible ways to improve the spatial resolution obtained for nominal parameters, 1.5 cm and 3.2 mm FWHM for human and small animal scanners, respectively, are indicated. Counting rates of true and random three-photon events for typical human and small animal scanning configurations are assessed. A simple formula for minimum size of lesions detectable in the three-gamma based images is derived. Depending on the contrast and total number of registered counts, lesions of a few mm size for human and sub mm for small animal scanners can be detected

Characterization of a cylindrical plastic β-detector with Monte Carlo simulations of optical photons

Energy Technology Data Exchange (ETDEWEB)

Guadilla, V., E-mail: victor.guadilla@ific.uv.es [Instituto de Física Corpuscular, CSIC-Universidad de Valencia, E-46071 Valencia (Spain); Algora, A. [Instituto de Física Corpuscular, CSIC-Universidad de Valencia, E-46071 Valencia (Spain); Institute of Nuclear Research of the Hungarian Academy of Sciences, Debrecen H-4026 (Hungary); Tain, J.L.; Agramunt, J. [Instituto de Física Corpuscular, CSIC-Universidad de Valencia, E-46071 Valencia (Spain); Äystö, J. [University of Jyvaskyla, Department of Physics, P.O. Box 35, FI-40014 (Finland); Briz, J.A.; Cucoanes, A. [Subatech, CNRS/IN2P3, Nantes, EMN, F-44307 Nantes (France); Eronen, T. [University of Jyvaskyla, Department of Physics, P.O. Box 35, FI-40014 (Finland); Estienne, M.; Fallot, M. [Subatech, CNRS/IN2P3, Nantes, EMN, F-44307 Nantes (France); Fraile, L.M. [Universidad Complutense, Grupo de Física Nuclear, CEI Moncloa, E-28040 Madrid (Spain); Ganioğlu, E. [Department of Physics, Istanbul University, 34134 Istanbul (Turkey); Gelletly, W. [Instituto de Física Corpuscular, CSIC-Universidad de Valencia, E-46071 Valencia (Spain); Department of Physics, University of Surrey, GU2 7XH Guildford (United Kingdom); Gorelov, D.; Hakala, J.; Jokinen, A. [University of Jyvaskyla, Department of Physics, P.O. Box 35, FI-40014 (Finland); Jordan, D. [Instituto de Física Corpuscular, CSIC-Universidad de Valencia, E-46071 Valencia (Spain); Kankainen, A.; Kolhinen, V.; Koponen, J. [University of Jyvaskyla, Department of Physics, P.O. Box 35, FI-40014 (Finland); and others

2017-05-11

In this work we report on the Monte Carlo study performed to understand and reproduce experimental measurements of a new plastic β-detector with cylindrical geometry. Since energy deposition simulations differ from the experimental measurements for such a geometry, we show how the simulation of production and transport of optical photons does allow one to obtain the shapes of the experimental spectra. Moreover, taking into account the computational effort associated with this kind of simulation, we develop a method to convert the simulations of energy deposited into light collected, depending only on the interaction point in the detector. This method represents a useful solution when extensive simulations have to be done, as in the case of the calculation of the response function of the spectrometer in a total absorption γ-ray spectroscopy analysis.

Accurate and efficient radiation transport in optically thick media -- by means of the Symbolic Implicit Monte Carlo method in the difference formulation

International Nuclear Information System (INIS)

Szoke, A; Brooks, E D; McKinley, M; Daffin, F

2005-01-01

The equations of radiation transport for thermal photons are notoriously difficult to solve in thick media without resorting to asymptotic approximations such as the diffusion limit. One source of this difficulty is that in thick, absorbing media thermal emission is almost completely balanced by strong absorption. In a previous publication [SB03], the photon transport equation was written in terms of the deviation of the specific intensity from the local equilibrium field. We called the new form of the equations the difference formulation. The difference formulation is rigorously equivalent to the original transport equation. It is particularly advantageous in thick media, where the radiation field approaches local equilibrium and the deviations from the Planck distribution are small. The difference formulation for photon transport also clarifies the diffusion limit. In this paper, the transport equation is solved by the Symbolic Implicit Monte Carlo (SIMC) method and a comparison is made between the standard formulation and the difference formulation. The SIMC method is easily adapted to the derivative source terms of the difference formulation, and a remarkable reduction in noise is obtained when the difference formulation is applied to problems involving thick media

Monte Carlo simulation of photon scattering in x-ray absorption imaging of high-intensity discharge lamps

Energy Technology Data Exchange (ETDEWEB)

Curry, J J, E-mail: jjcurry@nist.go [National Institute of Standards and Technology, Gaithersburg, MD 20899-8422 (United States)

2010-06-16

Coherent and incoherent scattering of x-rays during x-ray absorption imaging of high-intensity discharge lamps have been studied with Monte Carlo simulations developed specifically for this purpose. The Monte Carlo code is described and some initial results are discussed. Coherent scattering, because of its angular concentration in the forward direction, is found to be the most significant scattering mechanism. Incoherent scattering, although comparably strong, is not as significant because it results primarily in photons being scattered in the rearward direction and therefore out of the detector. Coherent scattering interferes with the detected absorption signal because the path of a scattered photon through the object to be imaged is unknown. Although scattering is usually a small effect, it can be significant in regions of high contrast. At the discharge/wall interface, as many as 50% of the detected photons are scattered photons. The effect of scattering on analysis of Hg distributions has not yet been quantified.

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

CERN Document Server

2002-01-01

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

Considerations of MCNP Monte Carlo code to be used as a radiotherapy treatment planning tool.

Science.gov (United States)

Juste, B; Miro, R; Gallardo, S; Verdu, G; Santos, A

2005-01-01

The present work has simulated the photon and electron transport in a Theratron 780® (MDS Nordion)60Co radiotherapy unit, using the Monte Carlo transport code, MCNP (Monte Carlo N-Particle). This project explains mainly the different methodologies carried out to speedup calculations in order to apply this code efficiently in radiotherapy treatment planning.

Monte Carlo electron-transport calculations for clinical beams using energy grouping

Energy Technology Data Exchange (ETDEWEB)

Teng, S P; Anderson, D W; Lindstrom, D G

1986-01-01

A Monte Carlo program has been utilized to study the penetration of broad electron beams into a water phantom. The MORSE-E code, originally developed for neutron and photon transport, was chosen for adaptation to electrons because of its versatility. The electron energy degradation model employed logarithmic spacing of electron energy groups and included effects of elastic scattering, inelastic-moderate-energy-loss-processes and inelastic-large-energy-loss-processes (catastrophic). Energy straggling and angular deflections were modeled from group to group, using the Moeller cross section for energy loss, and Goudsmit-Saunderson theory to describe angular deflections. The resulting energy- and electron-deposition distributions in depth were obtained at 10 and 20 MeV and are compared with ETRAN results and broad beam experimental data from clinical accelerators.

Two-dimensional discrete ordinates photon transport calculations for brachytherapy dosimetry applications

International Nuclear Information System (INIS)

Daskalov, G.M.; Baker, R.S.; Little, R.C.; Rogers, D.W.O.; Williamson, J.F.

2000-01-01

The DANTSYS discrete ordinates computer code system is applied to quantitative estimation of water kerma rate distributions in the vicinity of discrete photon sources with energies in the 20- to 800-keV range in two-dimensional cylindrical r-z geometry. Unencapsulated sources immersed in cylindrical water phantoms of 40-cm diameter and 40-cm height are modeled in either homogeneous phantoms or shielded by Ti, Fe, and Pb filters with thicknesses of 1 and 2 mean free paths. The obtained dose results are compared with corresponding photon Monte Carlo simulations. A 210-group photon cross-section library for applications in this energy range is developed and applied, together with a general-purpose 42-group library developed at Los Alamos National Laboratory, for DANTSYS calculations. The accuracy of DANTSYS with the 42-group library relative to Monte Carlo exhibits large pointwise fluctuations from -42 to +84%. The major cause for the observed discrepancies is determined to be the inadequacy of the weighting function used for the 42-group library derivation. DANTSYS simulations with a finer 210-group library show excellent accuracy on and off the source transverse plane relative to Monte Carlo kerma calculations, varying from minus4.9 to 3.7%. The P 3 Legendre polynomial expansion of the angular scattering function is shown to be sufficient for accurate calculations. The results demonstrate that DANTSYS is capable of calculating photon doses in very good agreement with Monte Carlo and that the multigroup cross-section library and efficient techniques for mitigation of ray effects are critical for accurate discrete ordinates implementation

Clinical dosimetry in photon radiotherapy. A Monte Carlo based investigation

International Nuclear Information System (INIS)

Wulff, Joerg

2010-01-01

Practical clinical dosimetry is a fundamental step within the radiation therapy process and aims at quantifying the absorbed radiation dose within a 1-2% uncertainty. To achieve this level of accuracy, corrections are needed for calibrated and air-filled ionization chambers, which are used for dose measurement. The procedures of correction are based on cavity theory of Spencer-Attix and are defined in current dosimetry protocols. Energy dependent corrections for deviations from calibration beams account for changed ionization chamber response in the treatment beam. The corrections applied are usually based on semi-analytical models or measurements and are generally hard to determine due to their magnitude of only a few percents or even less. Furthermore the corrections are defined for fixed geometrical reference-conditions and do not apply to non-reference conditions in modern radiotherapy applications. The stochastic Monte Carlo method for the simulation of radiation transport is becoming a valuable tool in the field of Medical Physics. As a suitable tool for calculation of these corrections with high accuracy the simulations enable the investigation of ionization chambers under various conditions. The aim of this work is the consistent investigation of ionization chamber dosimetry in photon radiation therapy with the use of Monte Carlo methods. Nowadays Monte Carlo systems exist, which enable the accurate calculation of ionization chamber response in principle. Still, their bare use for studies of this type is limited due to the long calculation times needed for a meaningful result with a small statistical uncertainty, inherent to every result of a Monte Carlo simulation. Besides heavy use of computer hardware, techniques methods of variance reduction to reduce the needed calculation time can be applied. Methods for increasing the efficiency in the results of simulation were developed and incorporated in a modern and established Monte Carlo simulation environment

New capabilities for Monte Carlo simulation of deuteron transport and secondary products generation

International Nuclear Information System (INIS)

Sauvan, P.; Sanz, J.; Ogando, F.

2010-01-01

Several important research programs are dedicated to the development of facilities based on deuteron accelerators. In designing these facilities, the definition of a validated computational approach able to simulate deuteron transport and evaluate deuteron interactions and production of secondary particles with acceptable precision is a very important issue. Current Monte Carlo codes, such as MCNPX or PHITS, when applied for deuteron transport calculations use built-in semi-analytical models to describe deuteron interactions. These models are found unreliable in predicting neutron and photon generated by low energy deuterons, typically present in those facilities. We present a new computational tool, resulting from an extension of the MCNPX code, which improve significantly the treatment of problems where any secondary product (neutrons, photons, tritons, etc.) generated by low energy deuterons reactions could play a major role. Firstly, it handles deuteron evaluated data libraries, which allow describing better low deuteron energy interactions. Secondly, it includes a reduction variance technique for production of secondary particles by charged particle-induced nuclear interactions, which allow reducing drastically the computing time needed in transport and nuclear response calculations. Verification of the computational tool is successfully achieved. This tool can be very helpful in addressing design issues such as selection of the dedicated neutron production target and accelerator radioprotection analysis. It can be also helpful to test the deuteron cross-sections under development in the frame of different international nuclear data programs.

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)

The numerical parallel computing of photon transport

International Nuclear Information System (INIS)

Huang Qingnan; Liang Xiaoguang; Zhang Lifa

1998-12-01

The parallel computing of photon transport is investigated, the parallel algorithm and the parallelization of programs on parallel computers both with shared memory and with distributed memory are discussed. By analyzing the inherent law of the mathematics and physics model of photon transport according to the structure feature of parallel computers, using the strategy of 'to divide and conquer', adjusting the algorithm structure of the program, dissolving the data relationship, finding parallel liable ingredients and creating large grain parallel subtasks, the sequential computing of photon transport into is efficiently transformed into parallel and vector computing. The program was run on various HP parallel computers such as the HY-1 (PVP), the Challenge (SMP) and the YH-3 (MPP) and very good parallel speedup has been gotten

An order αs Monte Carlo calculation of hadronic double photon production

International Nuclear Information System (INIS)

Owens, J.F.

1992-01-01

The results of an order α s calculation of hadronic double photon production are discussed and compared with data from both colliding beam and fixed target experiments. The calculation utilizes a combination of analytic and Monte Carlo integration methods which make it easy to calculate a variety of observables and impose experimental cuts. (author) 8 refs.; 2 figs

PyMercury: Interactive Python for the Mercury Monte Carlo Particle Transport Code

International Nuclear Information System (INIS)

Iandola, F.N.; O'Brien, M.J.; Procassini, R.J.

2010-01-01

Monte Carlo particle transport applications are often written in low-level languages (C/C++) for optimal performance on clusters and supercomputers. However, this development approach often sacrifices straightforward usability and testing in the interest of fast application performance. To improve usability, some high-performance computing applications employ mixed-language programming with high-level and low-level languages. In this study, we consider the benefits of incorporating an interactive Python interface into a Monte Carlo application. With PyMercury, a new Python extension to the Mercury general-purpose Monte Carlo particle transport code, we improve application usability without diminishing performance. In two case studies, we illustrate how PyMercury improves usability and simplifies testing and validation in a Monte Carlo application. In short, PyMercury demonstrates the value of interactive Python for Monte Carlo particle transport applications. In the future, we expect interactive Python to play an increasingly significant role in Monte Carlo usage and testing.

Modeling bioluminescent photon transport in tissue based on Radiosity-diffusion model

Science.gov (United States)

Sun, Li; Wang, Pu; Tian, Jie; Zhang, Bo; Han, Dong; Yang, Xin

2010-03-01

Bioluminescence tomography (BLT) is one of the most important non-invasive optical molecular imaging modalities. The model for the bioluminescent photon propagation plays a significant role in the bioluminescence tomography study. Due to the high computational efficiency, diffusion approximation (DA) is generally applied in the bioluminescence tomography. But the diffusion equation is valid only in highly scattering and weakly absorbing regions and fails in non-scattering or low-scattering tissues, such as a cyst in the breast, the cerebrospinal fluid (CSF) layer of the brain and synovial fluid layer in the joints. A hybrid Radiosity-diffusion model is proposed for dealing with the non-scattering regions within diffusing domains in this paper. This hybrid method incorporates a priori information of the geometry of non-scattering regions, which can be acquired by magnetic resonance imaging (MRI) or x-ray computed tomography (CT). Then the model is implemented using a finite element method (FEM) to ensure the high computational efficiency. Finally, we demonstrate that the method is comparable with Mont Carlo (MC) method which is regarded as a 'gold standard' for photon transportation simulation.

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

Science.gov (United States)

Foucart, Francois

2018-04-01

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

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

International Nuclear Information System (INIS)

Higuchi, Kenji; Takemiya, Hiroshi; Kawasaki, Takuji

1998-01-01

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

A Monte Carlo Green's function method for three-dimensional neutron transport

International Nuclear Information System (INIS)

Gamino, R.G.; Brown, F.B.; Mendelson, M.R.

1992-01-01

This paper describes a Monte Carlo transport kernel capability, which has recently been incorporated into the RACER continuous-energy Monte Carlo code. The kernels represent a Green's function method for neutron transport from a fixed-source volume out to a particular volume of interest. This method is very powerful transport technique. Also, since kernels are evaluated numerically by Monte Carlo, the problem geometry can be arbitrarily complex, yet exact. This method is intended for problems where an ex-core neutron response must be determined for a variety of reactor conditions. Two examples are ex-core neutron detector response and vessel critical weld fast flux. The response is expressed in terms of neutron transport kernels weighted by a core fission source distribution. In these types of calculations, the response must be computed for hundreds of source distributions, but the kernels only need to be calculated once. The advance described in this paper is that the kernels are generated with a highly accurate three-dimensional Monte Carlo transport calculation instead of an approximate method such as line-of-sight attenuation theory or a synthesized three-dimensional discrete ordinates solution

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

International Nuclear Information System (INIS)

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

1994-09-01

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

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

International Nuclear Information System (INIS)

1992-01-01

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

Radiation shielding techniques and applications. 4. Two-Phase Monte Carlo Approach to Photon Streaming Through Three-Legged Penetrations

International Nuclear Information System (INIS)

White, Travis; Hack, Joe; Nathan, Steve; Barnett, Marvin

2001-01-01

solutions for scattering of neutrons through multi-legged penetrations are readily available in the literature; similar analytical solutions for photon scattering through penetrations, however, are not. Therefore, computer modeling must be relied upon to perform our analyses. The computer code typically used by Westinghouse SMS in the evaluation of photon transport through complex geometries is the MCNP Monte Carlo computer code. Yet, geometries of this nature can cause problems even with the Monte Carlo codes. Striking a balance between how the code handles bulk transport through the wall with transport through the penetration void, particularly with the use of typical variance reduction methods, is difficult when trying to ensure that all the important regions of the model are sampled appropriately. The problem was broken down into several roughly independent cases. First, scatter through the penetration was considered. Second, bulk transport through the hot leg of the duct and then through the remaining thickness of wall was calculated to determine the amount of supplemental shielding required in the wall. Similar analyses were performed for the middle and cold legs of the penetration. Finally, additional external shielding from radiation streaming through the duct was determined for cases where the minimum offset distance was not feasible. Each case was broken down further into two phases. In the first phase of each case, photons were transported from the source material to an area at the face of the wall, or the opening of the duct, where photon energy and angular distributions were tallied, representing the source incident on the wall or opening. Then, a simplified model for each case was developed and analyzed using the data from the first phase and the new source term. (authors)

Cumulative percent energy deposition of photon beam incident on different targets, simulated by Monte Carlo

International Nuclear Information System (INIS)

Kandic, A.; Jevremovic, T.; Boreli, F.

1989-01-01

Monte Carlo simulation (without secondary radiation) of the standard photon interactions (Compton scattering, photoelectric absorption and pair protection) for the complex slab's geometry is used in numerical code ACCA. A typical ACCA run will yield: (a) transmission of primary photon radiation differential in energy, (b) the spectrum of energy deposited in the target as a function of position and (c) the cumulative percent energy deposition as a function of position. A cumulative percent energy deposition of photon monoenergetic beam incident on simplest and complexity tissue slab and Fe slab are presented in this paper. (author). 5 refs.; 2 figs

RCPO1 - A Monte Carlo program for solving neutron and photon transport problems in three dimensional geometry with detailed energy description and depletion capability

International Nuclear Information System (INIS)

Ondis, L.A. II; Tyburski, L.J.; Moskowitz, B.S.

2000-01-01

The RCP01 Monte Carlo program is used to analyze many geometries of interest in nuclear design and analysis of light water moderated reactors such as the core in its pressure vessel with complex piping arrangement, fuel storage arrays, shipping and container arrangements, and neutron detector configurations. Written in FORTRAN and in use on a variety of computers, it is capable of estimating steady state neutron or photon reaction rates and neutron multiplication factors. The energy range covered in neutron calculations is that relevant to the fission process and subsequent slowing-down and thermalization, i.e., 20 MeV to 0 eV. The same energy range is covered for photon calculations

RCPO1 - A Monte Carlo program for solving neutron and photon transport problems in three dimensional geometry with detailed energy description and depletion capability

Energy Technology Data Exchange (ETDEWEB)

Ondis, L.A., II; Tyburski, L.J.; Moskowitz, B.S.

2000-03-01

The RCP01 Monte Carlo program is used to analyze many geometries of interest in nuclear design and analysis of light water moderated reactors such as the core in its pressure vessel with complex piping arrangement, fuel storage arrays, shipping and container arrangements, and neutron detector configurations. Written in FORTRAN and in use on a variety of computers, it is capable of estimating steady state neutron or photon reaction rates and neutron multiplication factors. The energy range covered in neutron calculations is that relevant to the fission process and subsequent slowing-down and thermalization, i.e., 20 MeV to 0 eV. The same energy range is covered for photon calculations.

SRNA-2K5, Proton Transport Using 3-D by Monte Carlo Techniques

International Nuclear Information System (INIS)

Ilic, Radovan D.

2005-01-01

1 - Description of program or function: SRNA-2K5 performs Monte Carlo transport simulation of proton in 3D source and 3D geometry of arbitrary materials. The proton transport based on condensed history model, and on model of compound nuclei decays that creates in nonelastic nuclear interaction by proton absorption. 2 - Methods: The SRNA-2K5 package is developed for time independent simulation of proton transport by Monte Carlo techniques for numerical experiments in complex geometry, using PENGEOM from PENELOPE with different material compositions, and arbitrary spectrum of proton generated from the 3D source. This package developed for 3D proton dose distribution in proton therapy and dosimetry, and it was based on the theory of multiple scattering. The compound nuclei decay was simulated by our and Russian MSDM models using ICRU 49 and ICRU 63 data. If protons trajectory is divided on great number of steps, protons passage can be simulated according to Berger's Condensed Random Walk model. Conditions of angular distribution and fluctuation of energy loss determinate step length. Physical picture of these processes is described by stopping power, Moliere's angular distribution, Vavilov's distribution with Sulek's correction per all electron orbits, and Chadwick's cross sections for nonelastic nuclear interactions, obtained by his GNASH code. According to physical picture of protons passage and with probabilities of protons transition from previous to next stage, which is prepared by SRNADAT program, simulation of protons transport in all SRNA codes runs according to usual Monte Carlo scheme: (i) proton from the spectrum prepared for random choice of energy, position and space angle is emitted from the source; (ii) proton is loosing average energy on the step; (iii) on that step, proton experience a great number of collisions, and it changes direction of movement randomly chosen from angular distribution; (iv) random fluctuation is added to average energy loss; (v

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

The width of electron-photon cascades in air

International Nuclear Information System (INIS)

Allan, H.R.; Crannell, C.J.; Hough, J.H.; Shutie, P.F.; Sun, M.P.

1975-01-01

A new Monte-Carlo simulation of electron-photon cascade development is shown to give results in agreement with a) the calculations of Nordheim et al, b) simulations using the Berger-Seltzer electron photon transport programme and c) experimental measurements in water. However, the lateral spread is much narrower than that given by the Nishimura-Kamata functions or the Messel-Crawford tabulations. (orig.) [de

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

Science.gov (United States)

Smith, L. M.; Hochstedler, R. D.

1997-02-01

Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code).

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

International Nuclear Information System (INIS)

Smith, L.M.; Hochstedler, R.D.

1997-01-01

Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code)

Monte Carlo simulation for the estimation of iron in human whole ...

Indian Academy of Sciences (India)

2017-02-10

Feb 10, 2017 ... Monte Carlo N-particle (MCNP) code has been used to simulate the transport of gamma photon rays ... experimental data, and better than the theoretical XCOM values. ... tions in the materials, according to probability density.

A simplified spherical harmonic method for coupled electron-photon transport calculations

International Nuclear Information System (INIS)

Josef, J.A.

1996-12-01

In this thesis we have developed a simplified spherical harmonic method (SP N method) and associated efficient solution techniques for 2-D multigroup electron-photon transport calculations. The SP N method has never before been applied to charged-particle transport. We have performed a first time Fourier analysis of the source iteration scheme and the P 1 diffusion synthetic acceleration (DSA) scheme applied to the 2-D SP N equations. Our theoretical analyses indicate that the source iteration and P 1 DSA schemes are as effective for the 2-D SP N equations as for the 1-D S N equations. Previous analyses have indicated that the P 1 DSA scheme is unstable (with sufficiently forward-peaked scattering and sufficiently small absorption) for the 2-D S N equations, yet is very effective for the 1-D S N equations. In addition, we have applied an angular multigrid acceleration scheme, and computationally demonstrated that it performs as well for the 2-D SP N equations as for the 1-D S N equations. It has previously been shown for 1-D S N calculations that this scheme is much more effective than the DSA scheme when scattering is highly forward-peaked. We have investigated the applicability of the SP N approximation to two different physical classes of problems: satellite electronics shielding from geomagnetically trapped electrons, and electron beam problems. In the space shielding study, the SP N method produced solutions that are accurate within 10% of the benchmark Monte Carlo solutions, and often orders of magnitude faster than Monte Carlo. We have successfully modeled quasi-void problems and have obtained excellent agreement with Monte Carlo. We have observed that the SP N method appears to be too diffusive an approximation for beam problems. This result, however, is in agreement with theoretical expectations

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

International Nuclear Information System (INIS)

Liang Jingang; Xu Qi; Wang Kan; Liu Shiwen

2013-01-01

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

Scalable Domain Decomposed Monte Carlo Particle Transport

Energy Technology Data Exchange (ETDEWEB)

O' Brien, Matthew Joseph [Univ. of California, Davis, CA (United States)

2013-12-05

In this dissertation, we present the parallel algorithms necessary to run domain decomposed Monte Carlo particle transport on large numbers of processors (millions of processors). Previous algorithms were not scalable, and the parallel overhead became more computationally costly than the numerical simulation.

Specific absorbed fractions of electrons and photons for Rad-HUMAN phantom using Monte Carlo method

International Nuclear Information System (INIS)

Wang Wen; Hu Liqin; Cheng Mengyun; Long Pengcheng

2015-01-01

The specific absorbed fractions (SAF) for self- and cross-irradiation are effective tools for the internal dose estimation of inhalation and ingestion intakes of radionuclides. A set of SAFs of photons and electrons were calculated using the Rad-HUMAN phantom, which is a computational voxel phantom of a Chinese adult female that was created using the color photographic image of the Chinese Visible Human (CVH) data set by the FDS Team. The model can represent most Chinese adult female anatomical characteristics and can be taken as an individual phantom to investigate the difference of internal dose with Caucasians. In this study, the emission of mono-energetic photons and electrons of 10 keV to 4 MeV energy were calculated using the Monte Carlo particle transport calculation code MCNP. Results were compared with the values from ICRP reference and ORNL models. The results showed that SAF from the Rad-HUMAN have similar trends but are larger than those from the other two models. The differences were due to the racial and anatomical differences in organ mass and inter-organ distance. The SAFs based on the Rad-HUMAN phantom provide an accurate and reliable data for internal radiation dose calculations for Chinese females. (authors)

Parallel processing implementation for the coupled transport of photons and electrons using OpenMP

Science.gov (United States)

Doerner, Edgardo

2016-05-01

In this work the use of OpenMP to implement the parallel processing of the Monte Carlo (MC) simulation of the coupled transport for photons and electrons is presented. This implementation was carried out using a modified EGSnrc platform which enables the use of the Microsoft Visual Studio 2013 (VS2013) environment, together with the developing tools available in the Intel Parallel Studio XE 2015 (XE2015). The performance study of this new implementation was carried out in a desktop PC with a multi-core CPU, taking as a reference the performance of the original platform. The results were satisfactory, both in terms of scalability as parallelization efficiency.

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

Science.gov (United States)

Rodriguez, M.; Brualla, L.

2018-04-01

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

Monteray Mark-I: Computer program (PC-version) for shielding calculation with Monte Carlo method

International Nuclear Information System (INIS)

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

1998-01-01

A computer program for gamma ray shielding calculation using Monte Carlo method has been developed. The program is written in WATFOR77 language. The MONTERAY MARH-1 is originally developed by James Wood. The program was modified by the authors that the modified version is easily executed. Applying Monte Carlo method the program observe photon gamma transport in an infinity planar shielding with various thick. A photon gamma is observed till escape from the shielding or when its energy less than the cut off energy. Pair production process is treated as pure absorption process that annihilation photons generated in the process are neglected in the calculation. The out put data calculated by the program are total albedo, build-up factor, and photon spectra. The calculation result for build-up factor of a slab lead and water media with 6 MeV parallel beam gamma source shows that they are in agreement with published data. Hence the program is adequate as a shielding design tool for observing gamma radiation transport in various media

Successful vectorization - reactor physics Monte Carlo code

International Nuclear Information System (INIS)

Martin, W.R.

1989-01-01

Most particle transport Monte Carlo codes in use today are based on the ''history-based'' algorithm, wherein one particle history at a time is simulated. Unfortunately, the ''history-based'' approach (present in all Monte Carlo codes until recent years) is inherently scalar and cannot be vectorized. In particular, the history-based algorithm cannot take advantage of vector architectures, which characterize the largest and fastest computers at the current time, vector supercomputers such as the Cray X/MP or IBM 3090/600. However, substantial progress has been made in recent years in developing and implementing a vectorized Monte Carlo algorithm. This algorithm follows portions of many particle histories at the same time and forms the basis for all successful vectorized Monte Carlo codes that are in use today. This paper describes the basic vectorized algorithm along with descriptions of several variations that have been developed by different researchers for specific applications. These applications have been mainly in the areas of neutron transport in nuclear reactor and shielding analysis and photon transport in fusion plasmas. The relative merits of the various approach schemes will be discussed and the present status of known vectorization efforts will be summarized along with available timing results, including results from the successful vectorization of 3-D general geometry, continuous energy Monte Carlo. (orig.)

Photons in a partonic transport approach

Energy Technology Data Exchange (ETDEWEB)

Greif, Moritz; Senzel, Florian; Greiner, Carsten [Goethe Universitaet Frankfurt, Max-von-Laue-Str. 1 60438 Frankfurt am Main (Germany)

2015-07-01

Partonic transport approaches have proved to be valuable tools in describing the quark-gluon plasma, created in heavy-ion collisions. In this work, first steps towards a dynamical understanding of photonproduction in expanding heavy-ion collisions are presented. Several photon production processes are included in the partonic cascade BAMPS (Boltzmann Approach to Multi-Parton Scatterings). BAMPS provides a microscopic tool to study expanding fireballs, employing a stochastic method to solve the relativistic 3+1d Boltzmann equation. Subsequently, photon spectra can be investigated, and in particular, the influence of the quark-gluon plasma phase for the elliptic flow of photons is studied.

Cavity-photon-switched coherent transient transport in a double quantum waveguide

Energy Technology Data Exchange (ETDEWEB)

Abdullah, Nzar Rauf, E-mail: nra1@hi.is; Gudmundsson, Vidar, E-mail: vidar@raunvis.hi.is [Science Institute, University of Iceland, Dunhaga 3, IS-107 Reykjavik (Iceland); Tang, Chi-Shung [Department of Mechanical Engineering, National United University, 1, Lienda, 36003 Miaoli, Taiwan (China); Manolescu, Andrei [School of Science and Engineering, Reykjavik University, Menntavegur 1, IS-101 Reykjavik (Iceland)

2014-12-21

We study a cavity-photon-switched coherent electron transport in a symmetric double quantum waveguide. The waveguide system is weakly connected to two electron reservoirs, but strongly coupled to a single quantized photon cavity mode. A coupling window is placed between the waveguides to allow electron interference or inter-waveguide transport. The transient electron transport in the system is investigated using a quantum master equation. We present a cavity-photon tunable semiconductor quantum waveguide implementation of an inverter quantum gate, in which the output of the waveguide system may be selected via the selection of an appropriate photon number or “photon frequency” of the cavity. In addition, the importance of the photon polarization in the cavity, that is, either parallel or perpendicular to the direction of electron propagation in the waveguide system is demonstrated.

Monte Carlo simulation of spectrum changes in a photon beam due to a brass compensator

International Nuclear Information System (INIS)

Custidiano, E.R.; Valenzuela, M.R.; Dumont, J.L.; McDonnell, J.; Rene, L; Rodriguez Aguirre, J.M.

2011-01-01

Monte Carlo simulations were used to study the changes in the incident spectrum when a poly-energetic photon beam passes through a static brass compensator. The simulated photon beam spectrum was evaluated by comparing it against the incident spectra. We also discriminated the changes in the transmitted spectrum produced by each of the microscopic processes. (i.e. Rayleigh scattering, photoelectric effect, Compton scattering, and pair production). The results show that the relevant process in the energy range considered is the Compton Effect, as expected for composite materials of intermediate atomic number and energy range considered.

Monte Carlo simulation of spectrum changes in a photon beam due to a brass compensator

Energy Technology Data Exchange (ETDEWEB)

Custidiano, E.R., E-mail: ernesto7661@gmail.com [Department of Physics, FaCENA, UNNE, Av., Libertad 5470, C.P.3400, Corrientes (Argentina); Valenzuela, M.R., E-mail: meraqval@gmail.com [Department of Physics, FaCENA, UNNE, Av., Libertad 5470, C.P.3400, Corrientes (Argentina); Dumont, J.L., E-mail: Joseluis.Dumont@elekta.com [Elekta CMS Software, St.Louis, MO (United States); McDonnell, J., E-mail: josemc@express.com.ar [Cumbres Institute, Riobamba 1745, C.P.2000, Rosario, Santa Fe (Argentina); Rene, L, E-mail: luismrene@gmail.com [Radiotherapy Center, Crespo 953, C.P.2000, Rosario, Santa Fe (Argentina); Rodriguez Aguirre, J.M., E-mail: juakcho@gmail.com [Department of Physics, FaCENA, UNNE, Av., Libertad 5470, C.P.3400, Corrientes (Argentina)

2011-06-15

Monte Carlo simulations were used to study the changes in the incident spectrum when a poly-energetic photon beam passes through a static brass compensator. The simulated photon beam spectrum was evaluated by comparing it against the incident spectra. We also discriminated the changes in the transmitted spectrum produced by each of the microscopic processes. (i.e. Rayleigh scattering, photoelectric effect, Compton scattering, and pair production). The results show that the relevant process in the energy range considered is the Compton Effect, as expected for composite materials of intermediate atomic number and energy range considered.

Small photon beam measurements using radiochromic film and Monte Carlo simulations in a water phantom

International Nuclear Information System (INIS)

Garcia-Garduno, Olivia A.; Larraga-Gutierrez, Jose M.; Rodriguez-Villafuerte, Mercedes; Martinez-Davalos, Arnulfo; Celis, Miguel A.

2010-01-01

This work reports the use of both GafChromic EBT film immersed in a water phantom and Monte Carlo (MC) simulations for small photon beam stereotactic radiosurgery dosimetry. Circularly collimated photon beams with diameters in the 4-20 mm range of a dedicated 6 MV linear accelerator (Novalis (registered) , BrainLAB, Germany) were used to perform off-axis ratios, tissue maximum ratios and total scatter factors measurements, and MC simulations. GafChromic EBT film data show an excellent agreement with MC results (<2.7%) for all measured quantities.

Weighted-delta-tracking for Monte Carlo particle transport

International Nuclear Information System (INIS)

Morgan, L.W.G.; Kotlyar, D.

2015-01-01

Highlights: • This paper presents an alteration to the Monte Carlo Woodcock tracking technique. • The alteration improves computational efficiency within regions of high absorbers. • The rejection technique is replaced by a statistical weighting mechanism. • The modified Woodcock method is shown to be faster than standard Woodcock tracking. • The modified Woodcock method achieves a lower variance, given a specified accuracy. - Abstract: Monte Carlo particle transport (MCPT) codes are incredibly powerful and versatile tools to simulate particle behavior in a multitude of scenarios, such as core/criticality studies, radiation protection, shielding, medicine and fusion research to name just a small subset applications. However, MCPT codes can be very computationally expensive to run when the model geometry contains large attenuation depths and/or contains many components. This paper proposes a simple modification to the Woodcock tracking method used by some Monte Carlo particle transport codes. The Woodcock method utilizes the rejection method for sampling virtual collisions as a method to remove collision distance sampling at material boundaries. However, it suffers from poor computational efficiency when the sample acceptance rate is low. The proposed method removes rejection sampling from the Woodcock method in favor of a statistical weighting scheme, which improves the computational efficiency of a Monte Carlo particle tracking code. It is shown that the modified Woodcock method is less computationally expensive than standard ray-tracing and rejection-based Woodcock tracking methods and achieves a lower variance, given a specified accuracy

Collision of Physics and Software in the Monte Carlo Application Toolkit (MCATK)

Energy Technology Data Exchange (ETDEWEB)

Sweezy, Jeremy Ed [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-01-21

The topic is presented in a series of slides organized as follows: MCATK overview, development strategy, available algorithms, problem modeling (sources, geometry, data, tallies), parallelism, miscellaneous tools/features, example MCATK application, recent areas of research, and summary and future work. MCATK is a C++ component-based Monte Carlo neutron-gamma transport software library with continuous energy neutron and photon transport. Designed to build specialized applications and to provide new functionality in existing general-purpose Monte Carlo codes like MCNP, it reads ACE formatted nuclear data generated by NJOY. The motivation behind MCATK was to reduce costs. MCATK physics involves continuous energy neutron & gamma transport with multi-temperature treatment, static eigenvalue (k_eff and α) algorithms, time-dependent algorithm, and fission chain algorithms. MCATK geometry includes mesh geometries and solid body geometries. MCATK provides verified, unit-test Monte Carlo components, flexibility in Monte Carlo application development, and numerous tools such as geometry and cross section plotters.

Photon albedo coefficients as functions of μ/Zeff parameter

Directory of Open Access Journals (Sweden)

Ljubenov Vladan L.

2013-01-01

Full Text Available This paper presents the results of the analyses of photon reflection from planar targets for normal photon incidence and for different shielding materials (water, concrete, aluminum, iron, and copper, in the range of the initial photon energies from 20 keV to 300 keV. Calculations of photon reflection parameters based on the results of Monte Carlo simulations of the photon transport have been performed using MCNP4C code. Integral reflection coefficients, presented as functions of the ratio of total cross-section of photons and effective atomic number of target material, show universal behaviour for all the analyzed shielding materials in the selected energy domain.

Status of Monte Carlo at Los Alamos

International Nuclear Information System (INIS)

Thompson, W.L.; Cashwell, E.D.

1980-01-01

At Los Alamos the early work of Fermi, von Neumann, and Ulam has been developed and supplemented by many followers, notably Cashwell and Everett, and the main product today is the continuous-energy, general-purpose, generalized-geometry, time-dependent, coupled neutron-photon transport code called MCNP. The Los Alamos Monte Carlo research and development effort is concentrated in Group X-6. MCNP treats an arbitrary three-dimensional configuration of arbitrary materials in geometric cells bounded by first- and second-degree surfaces and some fourth-degree surfaces (elliptical tori). Monte Carlo has evolved into perhaps the main method for radiation transport calculations at Los Alamos. MCNP is used in every technical division at the Laboratory by over 130 users about 600 times a month accounting for nearly 200 hours of CDC-7600 time

RCP01: a Monte Carlo program for solving neutron and photon transport problems in three-dimensional geometry with detailed energy description (LWBR development program). [For CDC-6600 and -7600, in FORTRAN

Energy Technology Data Exchange (ETDEWEB)

Candelore, N R; Gast, R C; Ondis, II, L A

1978-08-01

The RCP01 Monte Carlo program for the CDC-7600 and CDC-6600 performs fixed source or eigenfunction neutron reaction rate calculations, or photon reaction rate calculations, for complex geometries. The photon calculations may be linked to the neutron reaction rate calculations. For neutron calculations, the full energy range is treated as required for neutron birth by the fission process and the subsequent neutron slowing down and thermalization, i.e., 10 MeV to 0 eV; for photon calculations the same energy range is treated. The detailed cross sections required for the neutron or photon collision processes are provided by RCPL1. This report provides details of the various types of neutron and photon starts and collisions, the common geometry tracking, and the input required. 37 figures, 1 table.

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

International Nuclear Information System (INIS)

Grieshemer, D.P.; Gill, D.F.; Nease, B.R.; Carpenter, D.C.; Joo, H.; Millman, D.L.; Sutton, T.M.; Stedry, M.H.; Dobreff, P.S.; Trumbull, T.H.; Caro, E.

2013-01-01

MC21 is a continuous-energy Monte Carlo radiation transport code for the calculation of the steady-state spatial distributions of reaction rates in three-dimensional models. The code supports neutron and photon transport in fixed source problems, as well as iterated-fission-source (eigenvalue) neutron transport problems. MC21 has been designed and optimized to support large-scale problems in reactor physics, shielding, and criticality analysis applications. The code also supports many in-line reactor feedback effects, including depletion, thermal feedback, xenon feedback, eigenvalue search, and neutron and photon heating. MC21 uses continuous-energy neutron/nucleus interaction physics over the range from 10 -5 eV to 20 MeV. The code treats all common neutron scattering mechanisms, including fast-range elastic and non-elastic scattering, and thermal- and epithermal-range scattering from molecules and crystalline materials. For photon transport, MC21 uses continuous-energy interaction physics over the energy range from 1 keV to 100 GeV. The code treats all common photon interaction mechanisms, including Compton scattering, pair production, and photoelectric interactions. All of the nuclear data required by MC21 is provided by the NDEX system of codes, which extracts and processes data from EPDL-, ENDF-, and ACE-formatted source files. For geometry representation, MC21 employs a flexible constructive solid geometry system that allows users to create spatial cells from first- and second-order surfaces. The system also allows models to be built up as hierarchical collections of previously defined spatial cells, with interior detail provided by grids and template overlays. Results are collected by a generalized tally capability which allows users to edit integral flux and reaction rate information. Results can be collected over the entire problem or within specific regions of interest through the use of phase filters that control which particles are allowed to score each

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

Monte Carlo method for neutron transport problems

International Nuclear Information System (INIS)

Asaoka, Takumi

1977-01-01

Some methods for decreasing variances in Monte Carlo neutron transport calculations are presented together with the results of sample calculations. A general purpose neutron transport Monte Carlo code ''MORSE'' was used for the purpose. The first method discussed in this report is the method of statistical estimation. As an example of this method, the application of the coarse-mesh rebalance acceleration method to the criticality calculation of a cylindrical fast reactor is presented. Effective multiplication factor and its standard deviation are presented as a function of the number of histories and comparisons are made between the coarse-mesh rebalance method and the standard method. Five-group neutron fluxes at core center are also compared with the result of S4 calculation. The second method is the method of correlated sampling. This method was applied to the perturbation calculation of control rod worths in a fast critical assembly (FCA-V-3) Two methods of sampling (similar flight paths and identical flight paths) are tested and compared with experimental results. For every cases the experimental value lies within the standard deviation of the Monte Carlo calculations. The third method is the importance sampling. In this report a biased selection of particle flight directions discussed. This method was applied to the flux calculation in a spherical fast neutron system surrounded by a 10.16 cm iron reflector. Result-direction biasing, path-length stretching, and no biasing are compared with S8 calculation. (Aoki, K.)

Soft Photons from transport and hydrodynamics at FAIR energies

International Nuclear Information System (INIS)

Grimm, Andreas; Bäuchle, Bjørn

2013-01-01

Direct photon spectra from uranium-uranium collisions at FAIR energies (E lab = 35 AGeV) are calculated within the hadronic Ultra-relativistic Quantum Molecular Dynamics transport model. In this microscopic model, one can optionally include a macroscopic intermediate hydrodynamic phase. The hot and dense stage of the collision is then modeled by a hydro dynamical calculation. Photon emission from transport-hydro hybrid calculations is examined for purely hadronic matter and matter that has a cross-over phase transition and a critical end point to deconfined and chirally restored matter at high temperatures. We find the photon spectra in both scenarios to be dominated by Bremsstrahlung. Comparing flow of photons in both cases suggests a way to distinguish these two scenarios.

Low-energy photons in high-energy photon fields--Monte Carlo generated spectra and a new descriptive parameter.

Science.gov (United States)

Chofor, Ndimofor; Harder, Dietrich; Willborn, Kay; Rühmann, Antje; Poppe, Björn

2011-09-01

The varying low-energy contribution to the photon spectra at points within and around radiotherapy photon fields is associated with variations in the responses of non-water equivalent dosimeters and in the water-to-material dose conversion factors for tissues such as the red bone marrow. In addition, the presence of low-energy photons in the photon spectrum enhances the RBE in general and in particular for the induction of second malignancies. The present study discusses the general rules valid for the low-energy spectral component of radiotherapeutic photon beams at points within and in the periphery of the treatment field, taking as an example the Siemens Primus linear accelerator at 6 MV and 15 MV. The photon spectra at these points and their typical variations due to the target system, attenuation, single and multiple Compton scattering, are described by the Monte Carlo method, using the code BEAMnrc/EGSnrc. A survey of the role of low energy photons in the spectra within and around radiotherapy fields is presented. In addition to the spectra, some data compression has proven useful to support the overview of the behaviour of the low-energy component. A characteristic indicator of the presence of low-energy photons is the dose fraction attributable to photons with energies not exceeding 200 keV, termed P(D)(200 keV). Its values are calculated for different depths and lateral positions within a water phantom. For a pencil beam of 6 or 15 MV primary photons in water, the radial distribution of P(D)(200 keV) is bellshaped, with a wide-ranging exponential tail of half value 6 to 7 cm. The P(D)(200 keV) value obtained on the central axis of a photon field shows an approximately proportional increase with field size. Out-of-field P(D)(200 keV) values are up to an order of magnitude higher than on the central axis for the same irradiation depth. The 2D pattern of P(D)(200 keV) for a radiotherapy field visualizes the regions, e.g. at the field margin, where changes of

Low dose out-of-field radiotherapy, part 2: Calculating the mean photon energy values for the out-of-field photon energy spectrum from scattered radiation using Monte Carlo methods.

Science.gov (United States)

Skrobala, A; Adamczyk, S; Kruszyna-Mochalska, M; Skórska, M; Konefał, A; Suchorska, W; Zaleska, K; Kowalik, A; Jackowiak, W; Malicki, J

2017-08-01

During radiotherapy, leakage from the machine head and collimator expose patients to out-of-field irradiation doses, which may cause secondary cancers. To quantify the risks of secondary cancers due to out-of-field doses, it is first necessary to measure these doses. Since most dosimeters are energy-dependent, it is essential to first determine the type of photon energy spectrum in the out-of-field area. The aim of this study was to determine the mean photon energy values for the out-of-field photon energy spectrum for a 6 MV photon beam using the GEANT 4-Monte Carlo method. A specially-designed large water phantom was simulated with a static field at gantry 0°. The source-to-surface distance was 92cm for an open field size of 10×10cm2. The photon energy spectra were calculated at five unique positions (at depths of 0.5, 1.6, 4, 6, 8, and 10cm) along the central beam axis and at six different off-axis distances. Monte Carlo simulations showed that mean radiation energy levels drop rapidly beyond the edge of the 6 MV photon beam field: at a distance of 10cm, the mean energy level is close to 0.3MeV versus 1.5MeV at the central beam axis. In some cases, the energy level actually increased even as the distance from the field edge increased: at a depth of 1.6cm and 15cm off-axis, the mean energy level was 0.205MeV versus 0.252MeV at 20cm off-axis. The out-of-field energy spectra and dose distribution data obtained in this study with Monte Carlo methods can be used to calibrate dosimeters to measure out-of-field radiation from 6MV photons. Copyright © 2017 Société française de radiothérapie oncologique (SFRO). Published by Elsevier SAS. All rights reserved.

Vectorizing and macrotasking Monte Carlo neutral particle algorithms

International Nuclear Information System (INIS)

Heifetz, D.B.

1987-04-01

Monte Carlo algorithms for computing neutral particle transport in plasmas have been vectorized and macrotasked. The techniques used are directly applicable to Monte Carlo calculations of neutron and photon transport, and Monte Carlo integration schemes in general. A highly vectorized code was achieved by calculating test flight trajectories in loops over arrays of flight data, isolating the conditional branches to as few a number of loops as possible. A number of solutions are discussed to the problem of gaps appearing in the arrays due to completed flights, which impede vectorization. A simple and effective implementation of macrotasking is achieved by dividing the calculation of the test flight profile among several processors. A tree of random numbers is used to ensure reproducible results. The additional memory required for each task may preclude using a larger number of tasks. In future machines, the limit of macrotasking may be possible, with each test flight, and split test flight, being a separate task

A computer code package for electron transport Monte Carlo simulation

International Nuclear Information System (INIS)

Popescu, Lucretiu M.

1999-01-01

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

Numerical radiation dosimetry using Monte Carlo photon transport at diagnostic energies

International Nuclear Information System (INIS)

Ioppolo, J.; Buckley, C.; Tuchyna, T.; Price, R.

2000-01-01

Full text: The Electron Gamma Shower 4 (EGS4) code has been installed on a WinNT workstation to allow the simulation of the dose absorbed in a patient during routine radiological examinations. Several additions to the code were required to form a theoretically sound model for use in the prediction of dose in the diagnostic energy range. Experimental measurements of dose using thermoluminescence dosimeters (TLD) were directly compared with EGS4 simulations. A Philips diagnostic X-ray machine with a field size of 17x17cm was used to irradiate a homogeneous perspex slab 30x30x12cm. TLDs were placed at evenly spaced points symmetrically about the central beam perpendicular to the cathode-anode axis at a number of depths. A diagnostic energy X-ray spectrum was measured from a comparable X-ray machine with similar beam quality and provided as input for the EGS4 code. Diverging point source geometry of the output beam, plus a realistic 3D model of the homogeneous perspex block, were also used in the EGS4 code. The LSCAT low energy photon scattering expansion by Namito et al (KEK internal report 95-10, 1995) was used to incorporate the binding effect that orbital electrons have on incoherent photons with energies less than 100 keV. EGS4 simulations were performed with sufficient numbers of photon histories to produce statistical uncertainties < 5% in the distribution of dose. Copyright (2000) Australasian College of Physical Scientists and Engineers in Medicine

Calibration and Monte Carlo modelling of neutron long counters

CERN Document Server

Tagziria, H

2000-01-01

The Monte Carlo technique has become a very powerful tool in radiation transport as full advantage is taken of enhanced cross-section data, more powerful computers and statistical techniques, together with better characterisation of neutron and photon source spectra. At the National Physical Laboratory, calculations using the Monte Carlo radiation transport code MCNP-4B have been combined with accurate measurements to characterise two long counters routinely used to standardise monoenergetic neutron fields. New and more accurate response function curves have been produced for both long counters. A novel approach using Monte Carlo methods has been developed, validated and used to model the response function of the counters and determine more accurately their effective centres, which have always been difficult to establish experimentally. Calculations and measurements agree well, especially for the De Pangher long counter for which details of the design and constructional material are well known. The sensitivit...

Environmental dose rate heterogeneity of beta radiation and its implications for luminescence dating: Monte Carlo modelling and experimental validation

DEFF Research Database (Denmark)

Nathan, R.P.; Thomas, P.J.; Jain, M.

2003-01-01

and identify the likely size of these effects on D-e distributions. The study employs the MCNP 4C Monte Carlo electron/photon transport model, supported by an experimental validation of the code in several case studies. We find good agreement between the experimental measurements and the Monte Carlo...

Monte Carlo impurity transport modeling in the DIII-D transport

International Nuclear Information System (INIS)

Evans, T.E.; Finkenthal, D.F.

1998-04-01

A description of the carbon transport and sputtering physics contained in the Monte Carlo Impurity (MCI) transport code is given. Examples of statistically significant carbon transport pathways are examined using MCI's unique tracking visualizer and a mechanism for enhanced carbon accumulation on the high field side of the divertor chamber is discussed. Comparisons between carbon emissions calculated with MCI and those measured in the DIII-D tokamak are described. Good qualitative agreement is found between 2D carbon emission patterns calculated with MCI and experimentally measured carbon patterns. While uncertainties in the sputtering physics, atomic data, and transport models have made quantitative comparisons with experiments more difficult, recent results using a physics based model for physical and chemical sputtering has yielded simulations with about 50% of the total carbon radiation measured in the divertor. These results and plans for future improvement in the physics models and atomic data are discussed

Transmutation approximations for the application of hybrid Monte Carlo/deterministic neutron transport to shutdown dose rate analysis

International Nuclear Information System (INIS)

Biondo, Elliott D.; Wilson, Paul P. H.

2017-01-01

In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using deterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation of an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed here, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR problem, GT-CADIS provides speedups of 200 100 relative to global variance reduction with the Forward-Weighted (FW)-CADIS method and 9 _± 5 • _1_0_"_4 relative to analog. As a result, this work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR analysis.

Monte Carlo simulations in skin radiotherapy

International Nuclear Information System (INIS)

Sarvari, A.; Jeraj, R.; Kron, T.

2000-01-01

The primary goal of this work was to develop a procedure for calculation the appropriate filter shape for a brachytherapy applicator used for skin radiotherapy. In the applicator a radioactive source is positioned close to the skin. Without a filter, the resultant dose distribution would be highly nonuniform.High uniformity is usually required however. This can be achieved using an appropriately shaped filter, which flattens the dose profile. Because of the complexity of the transport and geometry, Monte Carlo simulations had to be used. An 192 Ir high dose rate photon source was used. All necessary transport parameters were simulated with the MCNP4B Monte Carlo code. A highly efficient iterative procedure was developed, which enabled calculation of the optimal filter shape in only few iterations. The initially non-uniform dose distributions became uniform within a percent when applying the filter calculated by this procedure. (author)

A New Monte Carlo Neutron Transport Code at UNIST

International Nuclear Information System (INIS)

Lee, Hyunsuk; Kong, Chidong; Lee, Deokjung

2014-01-01

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

An Electron/Photon/Relaxation Data Library for MCNP6

Energy Technology Data Exchange (ETDEWEB)

Hughes, III, H. Grady [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-08-07

The capabilities of the MCNP6 Monte Carlo code in simulation of electron transport, photon transport, and atomic relaxation have recently been significantly expanded. The enhancements include not only the extension of existing data and methods to lower energies, but also the introduction of new categories of data and methods. Support of these new capabilities has required major additions to and redesign of the associated data tables. In this paper we present the first complete documentation of the contents and format of the new electron-photon-relaxation data library now available with the initial production release of MCNP6.

Inverse Monte Carlo: a unified reconstruction algorithm for SPECT

International Nuclear Information System (INIS)

Floyd, C.E.; Coleman, R.E.; Jaszczak, R.J.

1985-01-01

Inverse Monte Carlo (IMOC) is presented as a unified reconstruction algorithm for Emission Computed Tomography (ECT) providing simultaneous compensation for scatter, attenuation, and the variation of collimator resolution with depth. The technique of inverse Monte Carlo is used to find an inverse solution to the photon transport equation (an integral equation for photon flux from a specified source) for a parameterized source and specific boundary conditions. The system of linear equations so formed is solved to yield the source activity distribution for a set of acquired projections. For the studies presented here, the equations are solved using the EM (Maximum Likelihood) algorithm although other solution algorithms, such as Least Squares, could be employed. While the present results specifically consider the reconstruction of camera-based Single Photon Emission Computed Tomographic (SPECT) images, the technique is equally valid for Positron Emission Tomography (PET) if a Monte Carlo model of such a system is used. As a preliminary evaluation, experimentally acquired SPECT phantom studies for imaging Tc-99m (140 keV) are presented which demonstrate the quantitative compensation for scatter and attenuation for a two dimensional (single slice) reconstruction. The algorithm may be expanded in a straight forward manner to full three dimensional reconstruction including compensation for out of plane scatter

Periodically modulated single-photon transport in one-dimensional waveguide

Science.gov (United States)

Li, Xingmin; Wei, L. F.

2018-03-01

Single-photon transport along a one-dimension waveguide interacting with a quantum system (e.g., two-level atom) is a very useful and meaningful simplified model of the waveguide-based optical quantum devices. Thus, how to modulate the transport of the photons in the waveguide structures by adjusting certain external parameters should be particularly important. In this paper, we discuss how such a modulation could be implemented by periodically driving the energy splitting of the interacting atom and the atom-photon coupling strength. By generalizing the well developed time-independent full quantum mechanical theory in real space to the time-dependent one, we show that various sideband-transmission phenomena could be observed. This means that, with these modulations the photon has certain probabilities to transmit through the scattering atom in the other energy sidebands. Inversely, by controlling the sideband transmission the periodic modulations of the single photon waveguide devices could be designed for the future optical quantum information processing applications.

Single photon transport by a moving atom through sub-wavelength hole

International Nuclear Information System (INIS)

Afanasiev, A.E.; Melentiev, P.N.; Kuzin, A.A.; Kalatskiy, A.Yu.; Balykin, V.I.

2017-01-01

The results of investigation of photon transport through the subwavelength hole in the opaque screen by using single neutral atom are represented. The basis of the proposed and implemented method is the absorption of a photon by a neutral atom immediately before the subwavelength aperture, traveling of the atoms through the hole and emission of a photon on the other side of the screen. Realized method is the alternative approach to existing for photon transport through a subwavelength aperture: 1) self-sustained transmittance of a photon through the aperture according to the Bethe’s model; 2) extra ordinary transmission because of surface-plasmon excitation.

Angular biasing in implicit Monte-Carlo

International Nuclear Information System (INIS)

Zimmerman, G.B.

1994-01-01

Calculations of indirect drive Inertial Confinement Fusion target experiments require an integrated approach in which laser irradiation and radiation transport in the hohlraum are solved simultaneously with the symmetry, implosion and burn of the fuel capsule. The Implicit Monte Carlo method has proved to be a valuable tool for the two dimensional radiation transport within the hohlraum, but the impact of statistical noise on the symmetric implosion of the small fuel capsule is difficult to overcome. We present an angular biasing technique in which an increased number of low weight photons are directed at the imploding capsule. For typical parameters this reduces the required computer time for an integrated calculation by a factor of 10. An additional factor of 5 can also be achieved by directing even smaller weight photons at the polar regions of the capsule where small mass zones are most sensitive to statistical noise

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

Towards real-time photon Monte Carlo dose calculation in the cloud

Science.gov (United States)

Ziegenhein, Peter; Kozin, Igor N.; Kamerling, Cornelis Ph; Oelfke, Uwe

2017-06-01

Near real-time application of Monte Carlo (MC) dose calculation in clinic and research is hindered by the long computational runtimes of established software. Currently, fast MC software solutions are available utilising accelerators such as graphical processing units (GPUs) or clusters based on central processing units (CPUs). Both platforms are expensive in terms of purchase costs and maintenance and, in case of the GPU, provide only limited scalability. In this work we propose a cloud-based MC solution, which offers high scalability of accurate photon dose calculations. The MC simulations run on a private virtual supercomputer that is formed in the cloud. Computational resources can be provisioned dynamically at low cost without upfront investment in expensive hardware. A client-server software solution has been developed which controls the simulations and transports data to and from the cloud efficiently and securely. The client application integrates seamlessly into a treatment planning system. It runs the MC simulation workflow automatically and securely exchanges simulation data with the server side application that controls the virtual supercomputer. Advanced encryption standards were used to add an additional security layer, which encrypts and decrypts patient data on-the-fly at the processor register level. We could show that our cloud-based MC framework enables near real-time dose computation. It delivers excellent linear scaling for high-resolution datasets with absolute runtimes of 1.1 seconds to 10.9 seconds for simulating a clinical prostate and liver case up to 1% statistical uncertainty. The computation runtimes include the transportation of data to and from the cloud as well as process scheduling and synchronisation overhead. Cloud-based MC simulations offer a fast, affordable and easily accessible alternative for near real-time accurate dose calculations to currently used GPU or cluster solutions.

Monte Carlo simulation of photon buildup factors for shielding materials in diagnostic x-ray facilities

International Nuclear Information System (INIS)

Kharrati, Hedi; Agrebi, Amel; Karoui, Mohamed Karim

2012-01-01

Purpose: A simulation of buildup factors for ordinary concrete, steel, lead, plate glass, lead glass, and gypsum wallboard in broad beam geometry for photons energies from 10 keV to 150 keV at 5 keV intervals is presented. Methods: Monte Carlo N-particle radiation transport computer code has been used to determine the buildup factors for the studied shielding materials. Results: An example concretizing the use of the obtained buildup factors data in computing the broad beam transmission for tube potentials at 70, 100, 120, and 140 kVp is given. The half value layer, the tenth value layer, and the equilibrium tenth value layer are calculated from the broad beam transmission for these tube potentials. Conclusions: The obtained values compared with those calculated from the published data show the ability of these data to predict shielding transmission curves. Therefore, the buildup factors data can be combined with primary, scatter, and leakage x-ray spectra to provide a computationally based solution to broad beam transmission for barriers in shielding x-ray facilities.

Monte Carlo simulation of photon buildup factors for shielding materials in diagnostic x-ray facilities.

Science.gov (United States)

Kharrati, Hedi; Agrebi, Amel; Karoui, Mohamed Karim

2012-10-01

A simulation of buildup factors for ordinary concrete, steel, lead, plate glass, lead glass, and gypsum wallboard in broad beam geometry for photons energies from 10 keV to 150 keV at 5 keV intervals is presented. Monte Carlo N-particle radiation transport computer code has been used to determine the buildup factors for the studied shielding materials. An example concretizing the use of the obtained buildup factors data in computing the broad beam transmission for tube potentials at 70, 100, 120, and 140 kVp is given. The half value layer, the tenth value layer, and the equilibrium tenth value layer are calculated from the broad beam transmission for these tube potentials. The obtained values compared with those calculated from the published data show the ability of these data to predict shielding transmission curves. Therefore, the buildup factors data can be combined with primary, scatter, and leakage x-ray spectra to provide a computationally based solution to broad beam transmission for barriers in shielding x-ray facilities.

The calculation of dose from external photon exposures using reference human phantoms and Monte-Carlo methods. Pt. 1

International Nuclear Information System (INIS)

Kramer, R.; Zankl, M.; Williams, G.; Drexler, G.

1982-12-01

By the help of a Monte-Carlo program the dose that single organs, organ groups and bigger or smaller parts of body would receive on an average, caused by an irradiation definitely fixed by the geometry of irradiation and photon energy, can be determined. Thus the phantom in connection with the Monte-Carlo program can be used for several considerations as for example - calculation of dose from occupational exposures - calculation of dose from diagnostic procedures - calculation of dose from radiotherapy procedures. (orig.)

Visual Monte Carlo and its application to internal and external dosimetry

International Nuclear Information System (INIS)

Hunt, J.G.; Silva, F.C. da; Souza-Santos, D. de; Dantas, B.M.; Azeredo, A.; Malatova, I.; Foltanova, S.; Isakson, M.

2001-01-01

The program visual Monte Carlo (VMC), combined with voxel phantoms, and its application to three areas of radiation protection: calibration of in vivo measurement systems, dose calculations due to external sources of radiation, and the calculation of Specific Effective Energies is described in this paper. The simulation of photon transport through a voxel phantom requires a Monte Carlo program adapted to voxel geometries. VMC is written in Visual Basic trademark, a Microsoft Windows based program, which is easy to use and has an extensive graphic output. (orig.)

Commissioning of a medical accelerator photon beam Monte Carlo simulation using wide-field profiles

International Nuclear Information System (INIS)

Pena, J; Franco, L; Gomez, F; Iglesias, A; Lobato, R; Mosquera, J; Pazos, A; Pardo, J; Pombar, M; RodrIguez, A; Sendon, J

2004-01-01

A method for commissioning an EGSnrc Monte Carlo simulation of medical linac photon beams through wide-field lateral profiles at moderate depth in a water phantom is presented. Although depth-dose profiles are commonly used for nominal energy determination, our study shows that they are quite insensitive to energy changes below 0.3 MeV (0.6 MeV) for a 6 MV (15 MV) photon beam. Also, the depth-dose profile dependence on beam radius adds an additional uncertainty in their use for tuning nominal energy. Simulated 40 cm x 40 cm lateral profiles at 5 cm depth in a water phantom show greater sensitivity to both nominal energy and radius. Beam parameters could be determined by comparing only these curves with measured data

Commissioning of a medical accelerator photon beam Monte Carlo simulation using wide-field profiles

Energy Technology Data Exchange (ETDEWEB)

Pena, J [Departamento de Fisica de PartIculas, Facultade de Fisica, 15782 Santiago de Compostela (Spain); Franco, L [Departamento de Fisica de PartIculas, Facultade de Fisica, 15782 Santiago de Compostela (Spain); Gomez, F [Departamento de Fisica de PartIculas, Facultade de Fisica, 15782 Santiago de Compostela (Spain); Iglesias, A [Departamento de Fisica de PartIculas, Facultade de Fisica, 15782 Santiago de Compostela (Spain); Lobato, R [Hospital ClInico Universitario de Santiago, Santiago de Compostela (Spain); Mosquera, J [Hospital ClInico Universitario de Santiago, Santiago de Compostela (Spain); Pazos, A [Departamento de Fisica de PartIculas, Facultade de Fisica, 15782 Santiago de Compostela (Spain); Pardo, J [Departamento de Fisica de PartIculas, Facultade de Fisica, 15782 Santiago de Compostela (Spain); Pombar, M [Hospital ClInico Universitario de Santiago, Santiago de Compostela (Spain); RodrIguez, A [Departamento de Fisica de PartIculas, Facultade de Fisica, 15782 Santiago de Compostela (Spain); Sendon, J [Hospital ClInico Universitario de Santiago, Santiago de Compostela (Spain)

2004-11-07

A method for commissioning an EGSnrc Monte Carlo simulation of medical linac photon beams through wide-field lateral profiles at moderate depth in a water phantom is presented. Although depth-dose profiles are commonly used for nominal energy determination, our study shows that they are quite insensitive to energy changes below 0.3 MeV (0.6 MeV) for a 6 MV (15 MV) photon beam. Also, the depth-dose profile dependence on beam radius adds an additional uncertainty in their use for tuning nominal energy. Simulated 40 cm x 40 cm lateral profiles at 5 cm depth in a water phantom show greater sensitivity to both nominal energy and radius. Beam parameters could be determined by comparing only these curves with measured data.

Neutron contamination of Varian Clinac iX 10 MV photon beam using Monte Carlo simulation

International Nuclear Information System (INIS)

Yani, S; Haryanto, F; Arif, I; Tursinah, R; Rhani, M F; Soh, R C X

2016-01-01

High energy medical accelerators are commonly used in radiotherapy to increase the effectiveness of treatments. As we know neutrons can be emitted from a medical accelerator if there is an incident of X-ray that hits any of its materials. This issue becomes a point of view of many researchers. The neutron contamination has caused many problems such as image resolution and radiation protection for patients and radio oncologists. This study concerns the simulation of neutron contamination emitted from Varian Clinac iX 10 MV using Monte Carlo code system. As neutron production process is very complex, Monte Carlo simulation with MCNPX code system was carried out to study this contamination. The design of this medical accelerator was modelled based on the actual materials and geometry. The maximum energy of photons and neutron in the scoring plane was 10.5 and 2.239 MeV, respectively. The number and energy of the particles produced depend on the depth and distance from beam axis. From these results, it is pointed out that the neutron produced by linac 10 MV photon beam in a typical treatment is not negligible. (paper)

Tally and geometry definition influence on the computing time in radiotherapy treatment planning with MCNP Monte Carlo code.

Science.gov (United States)

Juste, B; Miro, R; Gallardo, S; Santos, A; Verdu, G

2006-01-01

The present work has simulated the photon and electron transport in a Theratron 780 (MDS Nordion) (60)Co radiotherapy unit, using the Monte Carlo transport code, MCNP (Monte Carlo N-Particle), version 5. In order to become computationally more efficient in view of taking part in the practical field of radiotherapy treatment planning, this work is focused mainly on the analysis of dose results and on the required computing time of different tallies applied in the model to speed up calculations.

Monte Carlo evaluation of a photon pencil kernel algorithm applied to fast neutron therapy treatment planning

Science.gov (United States)

Söderberg, Jonas; Alm Carlsson, Gudrun; Ahnesjö, Anders

2003-10-01

When dedicated software is lacking, treatment planning for fast neutron therapy is sometimes performed using dose calculation algorithms designed for photon beam therapy. In this work Monte Carlo derived neutron pencil kernels in water were parametrized using the photon dose algorithm implemented in the Nucletron TMS (treatment management system) treatment planning system. A rectangular fast-neutron fluence spectrum with energies 0-40 MeV (resembling a polyethylene filtered p(41)+ Be spectrum) was used. Central axis depth doses and lateral dose distributions were calculated and compared with the corresponding dose distributions from Monte Carlo calculations for homogeneous water and heterogeneous slab phantoms. All absorbed doses were normalized to the reference dose at 10 cm depth for a field of radius 5.6 cm in a 30 × 40 × 20 cm3 water test phantom. Agreement to within 7% was found in both the lateral and the depth dose distributions. The deviations could be explained as due to differences in size between the test phantom and that used in deriving the pencil kernel (radius 200 cm, thickness 50 cm). In the heterogeneous phantom, the TMS, with a directly applied neutron pencil kernel, and Monte Carlo calculated absorbed doses agree approximately for muscle but show large deviations for media such as adipose or bone. For the latter media, agreement was substantially improved by correcting the absorbed doses calculated in TMS with the neutron kerma factor ratio and the stopping power ratio between tissue and water. The multipurpose Monte Carlo code FLUKA was used both in calculating the pencil kernel and in direct calculations of absorbed dose in the phantom.

Cost of splitting in Monte Carlo transport

International Nuclear Information System (INIS)

Everett, C.J.; Cashwell, E.D.

1978-03-01

In a simple transport problem designed to estimate transmission through a plane slab of x free paths by Monte Carlo methods, it is shown that m-splitting (m > or = 2) does not pay unless exp(x) > m(m + 3)/(m - 1). In such a case, the minimum total cost in terms of machine time is obtained as a function of m, and the optimal value of m is determined

Techniques to reduce memory requirements for coupled photon-electron transport

International Nuclear Information System (INIS)

Turcksin, Bruno; Ragusa, Jean; Morel, Jim

2011-01-01

In this work, we present two methods to decrease memory needs while solving the photon- electron transport equation. The coupled transport of electrons and photons is of importance in radiotherapy because it describes the interactions of X-rays with matter. One of the issues of discretized electron transport is that the electron scattering is highly forward peaked. A common approximation is to represent the peak in the scattering cross section by a Dirac distribution. This is convenient, but the integration over all angles of this distribution requires the use of Galerkin quadratures. By construction these quadratures impose that the number of flux moments be equal to the number of directions (number of angular fluxes), which is very demanding in terms of memory. In this study, we show that even if the number of moments is not as large as the number of directions, an accurate solution can be obtained when using Galerkin quadratures. Another method to decrease the memory needs involves choosing an appropriate reordering of the energy groups. We show in this paper that an appropriate alternation of photons/electrons groups allows to rewrite one transport problem of n groups as gcd successive transport problems of n/gcd groups where gcd is the greatest common divisor between the number of photon groups and the number of electron groups. (author)

Efficient sampling algorithms for Monte Carlo based treatment planning

International Nuclear Information System (INIS)

DeMarco, J.J.; Solberg, T.D.; Chetty, I.; Smathers, J.B.

1998-01-01

Efficient sampling algorithms are necessary for producing a fast Monte Carlo based treatment planning code. This study evaluates several aspects of a photon-based tracking scheme and the effect of optimal sampling algorithms on the efficiency of the code. Four areas were tested: pseudo-random number generation, generalized sampling of a discrete distribution, sampling from the exponential distribution, and delta scattering as applied to photon transport through a heterogeneous simulation geometry. Generalized sampling of a discrete distribution using the cutpoint method can produce speedup gains of one order of magnitude versus conventional sequential sampling. Photon transport modifications based upon the delta scattering method were implemented and compared with a conventional boundary and collision checking algorithm. The delta scattering algorithm is faster by a factor of six versus the conventional algorithm for a boundary size of 5 mm within a heterogeneous geometry. A comparison of portable pseudo-random number algorithms and exponential sampling techniques is also discussed

Comparison of analytical and Monte Carlo calculations of multi-photon effects in bremsstrahlung emission by high-energy electrons

DEFF Research Database (Denmark)

Mangiarotti, Alessio; Sona, Pietro; Ballestrero, Sergio

2012-01-01

Approximate analytical calculations of multi-photon effects in the spectrum of total radiated energy by high-energy electrons crossing thin targets are compared to the results of Monte Carlo type simulations. The limits of validity of the analytical expressions found in the literature are establi...

Monte Carlo calculation of dose rate conversion factors for external exposure to photon emitters in soil

CERN Document Server

Clouvas, A; Antonopoulos-Domis, M; Silva, J

2000-01-01

The dose rate conversion factors D/sub CF/ (absorbed dose rate in air per unit activity per unit of soil mass, nGy h/sup -1/ per Bq kg/sup -1/) are calculated 1 m above ground for photon emitters of natural radionuclides uniformly distributed in the soil. Three Monte Carlo codes are used: 1) The MCNP code of Los Alamos; 2) The GEANT code of CERN; and 3) a Monte Carlo code developed in the Nuclear Technology Laboratory of the Aristotle University of Thessaloniki. The accuracy of the Monte Carlo results is tested by the comparison of the unscattered flux obtained by the three Monte Carlo codes with an independent straightforward calculation. All codes and particularly the MCNP calculate accurately the absorbed dose rate in air due to the unscattered radiation. For the total radiation (unscattered plus scattered) the D/sub CF/ values calculated from the three codes are in very good agreement between them. The comparison between these results and the results deduced previously by other authors indicates a good ag...

Nuclear data physics issues in Monte Carlo simulations of neutron and photon transport in the Indian context

International Nuclear Information System (INIS)

Ganesan, S.

2009-01-01

In this write-up, some of the basic issues of nuclear data physics in Monte Carlo simulation of neutron transport in the Indian context are dealt with. In this lecture, some of the aspects associated with usage of the ENDF/B system, and of the PREPRO code system developed by D.E. Cullen and distributed by the IAEA Nuclear Data Section are briefly touched upon. Some aspects of the SIGACE code system which was developed by the author in collaboration with IPR, Ahmedabad and the IAEA Nuclear Data Section are also briefly covered. The validation of the SIGACE package included investigations using the NJOY and the MCNP compatible ACE files. Appendix-1 of the paper provides some useful discussions pointing out that voluminous and high-quality nuclear physics data required for nuclear applications usually evolve from a national effort to provide state-of-the-art data that are based upon established needs and uncertainties. Appendix-2 deals with some interesting work that was carried out using the SIGACE Code for Generating High Temperature ACE Files. Appendix-3 mentions briefly Integral nuclear data validation studies and use of Monte Carlo codes and nuclear data. Appendix-4 provides a brief summary report on selected Indian nuclear data physics activities for the interested reader in the light of BARC/DAE treating the subject area of nuclear data physics as a thrust area in our atomic energy programme

Monte Carlo-based investigations on the impact of removing the flattening filter on beam quality specifiers for photon beam dosimetry.

Science.gov (United States)

Czarnecki, Damian; Poppe, Björn; Zink, Klemens

2017-06-01

The impact of removing the flattening filter in clinical electron accelerators on the relationship between dosimetric quantities such as beam quality specifiers and the mean photon and electron energies of the photon radiation field was investigated by Monte Carlo simulations. The purpose of this work was to determine the uncertainties when using the well-known beam quality specifiers or energy-based beam specifiers as predictors of dosimetric photon field properties when removing the flattening filter. Monte Carlo simulations applying eight different linear accelerator head models with and without flattening filter were performed in order to generate realistic radiation sources and calculate field properties such as restricted mass collision stopping power ratios (L¯/ρ)airwater, mean photon and secondary electron energies. To study the impact of removing the flattening filter on the beam quality correction factors k Q , this factor for detailed ionization chamber models was calculated by Monte Carlo simulations. Stopping power ratios (L¯/ρ)airwater and k Q values for different ionization chambers as a function of TPR1020 and %dd(10) x were calculated. Moreover, mean photon energies in air and at the point of measurement in water as well as mean secondary electron energies at the point of measurement were calculated. The results revealed that removing the flattening filter led to a change within 0.3% in the relationship between %dd(10) x and (L¯/ρ)airwater, whereby the relationship between TPR1020 and (L¯/ρ)airwater changed up to 0.8% for high energy photon beams. However, TPR1020 was a good predictor of (L¯/ρ)airwater for both types of linear accelerator with energies filter within 1.1% and 1.6% was observed for TPR1020 and %dd(10) x respectively. The results of this study have shown that removing the flattening filter led to a change in the relationship between the well-known beam quality specifiers and dosimetric quantities at the point of measurement

Simulation of neutron transport equation using parallel Monte Carlo for deep penetration problems

International Nuclear Information System (INIS)

Bekar, K. K.; Tombakoglu, M.; Soekmen, C. N.

2001-01-01

Neutron transport equation is simulated using parallel Monte Carlo method for deep penetration neutron transport problem. Monte Carlo simulation is parallelized by using three different techniques; direct parallelization, domain decomposition and domain decomposition with load balancing, which are used with PVM (Parallel Virtual Machine) software on LAN (Local Area Network). The results of parallel simulation are given for various model problems. The performances of the parallelization techniques are compared with each other. Moreover, the effects of variance reduction techniques on parallelization are discussed

MCSLTT, Monte Carlo Simulation of Light Transport in Tissue

International Nuclear Information System (INIS)

2008-01-01

Description of program or function: Understanding light-tissue interaction is fundamental in the field of Biomedical Optics. It has important implications for both therapeutic and diagnostic technologies. In this program, light transport in scattering tissue is modeled by absorption and scattering events as each photon travels through the tissue. The path of each photon is determined statistically by calculating probabilities of scattering and absorption. Other measured quantities are total reflected light, total transmitted light, and total heat absorbed

Monte Carlo method for neutron transport calculations in graphics processing units (GPUs)

International Nuclear Information System (INIS)

Pellegrino, Esteban

2011-01-01

Monte Carlo simulation is well suited for solving the Boltzmann neutron transport equation in an inhomogeneous media for complicated geometries. However, routine applications require the computation time to be reduced to hours and even minutes in a desktop PC. The interest in adopting Graphics Processing Units (GPUs) for Monte Carlo acceleration is rapidly growing. This is due to the massive parallelism provided by the latest GPU technologies which is the most promising solution to the challenge of performing full-size reactor core analysis on a routine basis. In this study, Monte Carlo codes for a fixed-source neutron transport problem were developed for GPU environments in order to evaluate issues associated with computational speedup using GPUs. Results obtained in this work suggest that a speedup of several orders of magnitude is possible using the state-of-the-art GPU technologies. (author) [es

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

FENDL/MG-2.0 and FENDL/MC-2.0. The processed cross-section libraries for neutron photon transport calculations. Version 1, March 1997. Summary documentation

International Nuclear Information System (INIS)

Wienke, H.; Herman, M.

1998-01-01

Evaluated neutron reaction data and photon-atom interaction cross sections for materials contained in the general purpose Fusion Evaluated Nuclear Data Library (FENDL/E2.0) have been processed with the NJOY code system into VITAMIN-J multigroup structure, for use in discrete-ordinates transport codes, and into continuous energy ACE format, for use in the Monte Carlo transport code MCNP. This document summarizes the resulting data libraries FENDL/MG-2.0 version 1 and FENDL/MC-2.0 version 1. The data are available costfree from the IAEA Nuclear Data Section online or on magnetic tape. (author)

Computational methods of electron/photon transport

International Nuclear Information System (INIS)

Mack, J.M.

1983-01-01

A review of computational methods simulating the non-plasma transport of electrons and their attendant cascades is presented. Remarks are mainly restricted to linearized formalisms at electron energies above 1 keV. The effectiveness of various metods is discussed including moments, point-kernel, invariant imbedding, discrete-ordinates, and Monte Carlo. Future research directions and the potential impact on various aspects of science and engineering are indicated

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

International Nuclear Information System (INIS)

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

2013-01-01

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

A user's guide to MICAP: A Monte Carlo Ionization Chamber Analysis Package

Energy Technology Data Exchange (ETDEWEB)

Johnson, J.O.; Gabriel, T.A.

1988-01-01

A collection of computer codes entitled MICAP - A Monte Carlo Ionization Chamber Analysis Package has been developed to determine the response of a gas-filled cavity ionization chamber in a mixed neutron and photon radiation environment. In particular, MICAP determines the neutron, photon, and total response of the ionization chamber. The applicability of MICAP encompasses all aspects of mixed field dosimetry analysis including detector design, preexperimental planning and post-experimental analysis. The MICAP codes include: RDNDF for reading and processing ENDF/B-formatted cross section files, MICRO for manipulating microscopic cross section data sets, MACRO for creating macroscopic cross section data sets, NEUTRON for transporting neutrons, RECOMB for calculating correction data due to ionization chamber saturation effects, HEAVY for transporting recoil heavy ions and charged particles, PECSP for generating photon and electron cross section and material data sets, PHOTPREP for generating photon source input tapes, and PHOTON for transporting photons and electrons. The codes are generally tailored to provide numerous input options, but whenever possible, default values are supplied which yield adequate results. All of the MICAP codes function independently, and are operational on the ORNL IBM 3033 computer system. 14 refs., 27 figs., 49 tabs.

Monte Carlo benchmark calculations of energy deposition by electron/photon showers up to 1 GeV

International Nuclear Information System (INIS)

Mehlhorn, T.A.; Halbleib, J.A.

1983-01-01

Over the past several years the TIGER series of coupled electron/photon Monte Carlo transport codes has been applied to a variety of problems involving nuclear and space radiations, electron accelerators, and radioactive sources. In particular, they have been used at Sandia to simulate the interaction of electron beams, generated by pulsed-power accelerators, with various target materials for weapons effect simulation, and electron beam fusion. These codes are based on the ETRAN system which was developed for an energy range from about 10 keV up to a few tens of MeV. In this paper we will discuss the modifications that were made to the TIGER series of codes in order to extend their applicability to energies of interest to the high energy physics community (up to 1 GeV). We report the results of a series of benchmark calculations of the energy deposition by high energy electron beams in various materials using the modified codes. These results are then compared with the published results of various experimental measurements and other computational models

Monte Carlo modeling of a High-Sensitivity MOSFET dosimeter for low- and medium-energy photon sources

International Nuclear Information System (INIS)

Wang, Brian; Kim, C.-H.; Xu, X. George

2004-01-01

Metal-oxide-semiconductor field effect transistor (MOSFET) dosimeters are increasingly utilized in radiation therapy and diagnostic radiology. While it is difficult to characterize the dosimeter responses for monoenergetic sources by experiments, this paper reports a detailed Monte Carlo simulation model of the High-Sensitivity MOSFET dosimeter using Monte Carlo N-Particle (MCNP) 4C. A dose estimator method was used to calculate the dose in the extremely thin sensitive volume. Efforts were made to validate the MCNP model using three experiments: (1) comparison of the simulated dose with the measurement of a Cs-137 source, (2) comparison of the simulated dose with analytical values, and (3) comparison of the simulated energy dependence with theoretical values. Our simulation results show that the MOSFET dosimeter has a maximum response at about 40 keV of photon energy. The energy dependence curve is also found to agree with the predicted value from theory within statistical uncertainties. The angular dependence study shows that the MOSFET dosimeter has a higher response (about 8%) when photons come from the epoxy side, compared with the kapton side for the Cs-137 source

The penetration, diffusion and energy deposition of high-energy photon in layered media

International Nuclear Information System (INIS)

Zhengming, Luo; Chengjun, Gou; Laub, Wolfram

2002-01-01

This paper presents a new theory for calculating the transport of high-energy photons and their secondary charged particles. We call this new algorithm characteristic line method, which is completely analytic. Using this new method we can not only accurately calculate the transport behavior of energetic photons, but also precisely describes the transport behavior and energy deposition of secondary electrons, photoelectrons, Compton recoil electrons and positron-electron pairs. Its calculation efficiency is much higher than the Monte Carlo method's. The theory can be directly applied to layered media situation and obtain a pencil-beam-modeled solution. Therefore, it may be applied to clinical applications for radiation therapy

Adaptable three-dimensional Monte Carlo modeling of imaged blood vessels in skin

Science.gov (United States)

Pfefer, T. Joshua; Barton, Jennifer K.; Chan, Eric K.; Ducros, Mathieu G.; Sorg, Brian S.; Milner, Thomas E.; Nelson, J. Stuart; Welch, Ashley J.

1997-06-01

In order to reach a higher level of accuracy in simulation of port wine stain treatment, we propose to discard the typical layered geometry and cylindrical blood vessel assumptions made in optical models and use imaging techniques to define actual tissue geometry. Two main additions to the typical 3D, weighted photon, variable step size Monte Carlo routine were necessary to achieve this goal. First, optical low coherence reflectometry (OLCR) images of rat skin were used to specify a 3D material array, with each entry assigned a label to represent the type of tissue in that particular voxel. Second, the Monte Carlo algorithm was altered so that when a photon crosses into a new voxel, the remaining path length is recalculated using the new optical properties, as specified by the material array. The model has shown good agreement with data from the literature. Monte Carlo simulations using OLCR images of asymmetrically curved blood vessels show various effects such as shading, scattering-induced peaks at vessel surfaces, and directionality-induced gradients in energy deposition. In conclusion, this augmentation of the Monte Carlo method can accurately simulate light transport for a wide variety of nonhomogeneous tissue geometries.

Monte Carlo Numerical Models for Nuclear Logging Applications

Directory of Open Access Journals (Sweden)

Fusheng Li

2012-06-01

Full Text Available Nuclear logging is one of most important logging services provided by many oil service companies. The main parameters of interest are formation porosity, bulk density, and natural radiation. Other services are also provided from using complex nuclear logging tools, such as formation lithology/mineralogy, etc. Some parameters can be measured by using neutron logging tools and some can only be measured by using a gamma ray tool. To understand the response of nuclear logging tools, the neutron transport/diffusion theory and photon diffusion theory are needed. Unfortunately, for most cases there are no analytical answers if complex tool geometry is involved. For many years, Monte Carlo numerical models have been used by nuclear scientists in the well logging industry to address these challenges. The models have been widely employed in the optimization of nuclear logging tool design, and the development of interpretation methods for nuclear logs. They have also been used to predict the response of nuclear logging systems for forward simulation problems. In this case, the system parameters including geometry, materials and nuclear sources, etc., are pre-defined and the transportation and interactions of nuclear particles (such as neutrons, photons and/or electrons in the regions of interest are simulated according to detailed nuclear physics theory and their nuclear cross-section data (probability of interacting. Then the deposited energies of particles entering the detectors are recorded and tallied and the tool responses to such a scenario are generated. A general-purpose code named Monte Carlo N– Particle (MCNP has been the industry-standard for some time. In this paper, we briefly introduce the fundamental principles of Monte Carlo numerical modeling and review the physics of MCNP. Some of the latest developments of Monte Carlo Models are also reviewed. A variety of examples are presented to illustrate the uses of Monte Carlo numerical models

Recent developments in discrete ordinates electron transport

International Nuclear Information System (INIS)

Morel, J.E.; Lorence, L.J. Jr.

1986-01-01

The discrete ordinates method is a deterministic method for numerically solving the Boltzmann equation. It was originally developed for neutron transport calculations, but is routinely used for photon and coupled neutron-photon transport calculations as well. The computational state of the art for coupled electron-photon transport (CEPT) calculations is not as developed as that for neutron transport calculations. The only production codes currently available for CEPT calculations are condensed-history Monte Carlo codes such as the ETRAN and ITS codes. A deterministic capability for production calculations is clearly needed. In response to this need, we have begun the development of a production discrete ordinates code for CEPT calculations. The purpose of this paper is to describe the basic approach we are taking, discuss the current status of the project, and present some new computational results. Although further characterization of the coupled electron-photon discrete ordinates method remains to be done, the results to date indicate that the discrete ordinates method can be just as accurate and from 10 to 100 times faster than the Monte Carlo method for a wide variety of problems. We stress that these results are obtained with standard discrete ordinates codes such as ONETRAN. It is clear that even greater efficiency can be obtained by developing a new generation of production discrete ordinates codes specifically designed to solve the Boltzmann-Fokker-Planck equation. However, the prospects for such development in the near future appear to be remote

Chain segmentation for the Monte Carlo solution of particle transport problems

International Nuclear Information System (INIS)

Ragheb, M.M.H.

1984-01-01

A Monte Carlo approach is proposed where the random walk chains generated in particle transport simulations are segmented. Forward and adjoint-mode estimators are then used in conjunction with the firstevent source density on the segmented chains to obtain multiple estimates of the individual terms of the Neumann series solution at each collision point. The solution is then constructed by summation of the series. The approach is compared to the exact analytical and to the Monte Carlo nonabsorption weighting method results for two representative slowing down and deep penetration problems. Application of the proposed approach leads to unbiased estimates for limited numbers of particle simulations and is useful in suppressing an effective bias problem observed in some cases of deep penetration particle transport problems

Monte Carlo radiation transport: A revolution in science

International Nuclear Information System (INIS)

Hendricks, J.

1993-01-01

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

Application of a Java-based, univel geometry, neutral particle Monte Carlo code to the searchlight problem

International Nuclear Information System (INIS)

Charles A. Wemple; Joshua J. Cogliati

2005-01-01

A univel geometry, neutral particle Monte Carlo transport code, written entirely in the Java programming language, is under development for medical radiotherapy applications. The code uses ENDF-VI based continuous energy cross section data in a flexible XML format. Full neutron-photon coupling, including detailed photon production and photonuclear reactions, is included. Charged particle equilibrium is assumed within the patient model so that detailed transport of electrons produced by photon interactions may be neglected. External beam and internal distributed source descriptions for mixed neutron-photon sources are allowed. Flux and dose tallies are performed on a univel basis. A four-tap, shift-register-sequence random number generator is used. Initial verification and validation testing of the basic neutron transport routines is underway. The searchlight problem was chosen as a suitable first application because of the simplicity of the physical model. Results show excellent agreement with analytic solutions. Computation times for similar numbers of histories are comparable to other neutron MC codes written in C and FORTRAN

Comparative Criticality Analysis of Two Monte Carlo Codes on Centrifugal Atomizer: MCNPS and SCALE

International Nuclear Information System (INIS)

Kang, H-S; Jang, M-S; Kim, S-R; Park, J-M; Kim, K-N

2015-01-01

There are two well-known Monte Carlo codes for criticality analysis, MCNP5 and SCALE. MCNP5 is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical system as a main analysis code. SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. SCALE was conceived and funded by US NRC to perform standardized computer analysis for licensing evaluation and is used widely in the world. We performed a validation test of MCNP5 and a comparative analysis of Monte Carlo codes, MCNP5 and SCALE, in terms of the critical analysis of centrifugal atomizer. In the criticality analysis using MCNP5 code, we obtained the statistically reliable results by using a large number of source histories per cycle and performing of uncertainty analysis

Comparative Criticality Analysis of Two Monte Carlo Codes on Centrifugal Atomizer: MCNPS and SCALE

Energy Technology Data Exchange (ETDEWEB)

Kang, H-S; Jang, M-S; Kim, S-R [NESS, Daejeon (Korea, Republic of); Park, J-M; Kim, K-N [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

There are two well-known Monte Carlo codes for criticality analysis, MCNP5 and SCALE. MCNP5 is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical system as a main analysis code. SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. SCALE was conceived and funded by US NRC to perform standardized computer analysis for licensing evaluation and is used widely in the world. We performed a validation test of MCNP5 and a comparative analysis of Monte Carlo codes, MCNP5 and SCALE, in terms of the critical analysis of centrifugal atomizer. In the criticality analysis using MCNP5 code, we obtained the statistically reliable results by using a large number of source histories per cycle and performing of uncertainty analysis.

Design and commissioning of the photon monitors and optical transport lines for the advanced photon source positron accumulator ring

International Nuclear Information System (INIS)

Berg, W.; Yang, B.; Lumpkin, A.; Jones, J.

1996-01-01

Two photon monitors have been designed and installed in the positron accumulator ring (PAR) of the Advanced Photon Source. The photon monitors characterize the beam's transverse profile, bunch length, emittance, and energy spread in a nonintrusive manner. An optical transport line delivers synchrotron light from the PAR out of a high radiation environment. Both charge-coupled device and fast-gated, intensified cameras are used to measure the transverse beam profile (0.11 - 1 mm for damped beam) with a resolution of 0.06 mm. A streak camera (θ τ =I ps) is used to measure the bunch length which is in the range of 0.3-1 ns. The design of the various transport components and commissioning results of the photon monitors will be discussed

Development of general-purpose particle and heavy ion transport monte carlo code

International Nuclear Information System (INIS)

Iwase, Hiroshi; Nakamura, Takashi; Niita, Koji

2002-01-01

The high-energy particle transport code NMTC/JAM, which has been developed at JAERI, was improved for the high-energy heavy ion transport calculation by incorporating the JQMD code, the SPAR code and the Shen formula. The new NMTC/JAM named PHITS (Particle and Heavy-Ion Transport code System) is the first general-purpose heavy ion transport Monte Carlo code over the incident energies from several MeV/nucleon to several GeV/nucleon. (author)

Application of Macro Response Monte Carlo method for electron spectrum simulation

International Nuclear Information System (INIS)

Perles, L.A.; Almeida, A. de

2007-01-01

During the past years several variance reduction techniques for Monte Carlo electron transport have been developed in order to reduce the electron computation time transport for absorbed dose distribution. We have implemented the Macro Response Monte Carlo (MRMC) method to evaluate the electron spectrum which can be used as a phase space input for others simulation programs. Such technique uses probability distributions for electron histories previously simulated in spheres (called kugels). These probabilities are used to sample the primary electron final state, as well as the creation secondary electrons and photons. We have compared the MRMC electron spectra simulated in homogeneous phantom against the Geant4 spectra. The results showed an agreement better than 6% in the spectra peak energies and that MRMC code is up to 12 time faster than Geant4 simulations

SU-F-T-370: A Fast Monte Carlo Dose Engine for Gamma Knife

Energy Technology Data Exchange (ETDEWEB)

Song, T; Zhou, L [Southern Medical University, Guangzhou, Guangdong (China); Li, Y [Beihang University, Beijing, Beijing (China)

2016-06-15

Purpose: To develop a fast Monte Carlo dose calculation algorithm for Gamma Knife. Methods: To make the simulation more efficient, we implemented the track repeating technique on GPU. We first use EGSnrc to pre-calculate the photon and secondary electron tracks in water from two mono-energy photons of 60Co. The total photon mean free paths for different materials and energies are obtained from NIST. During simulation, each entire photon track was first loaded to shared memory for each block, the incident original photon was then splitted to Nthread sub-photons, each thread transport one sub-photon, the Russian roulette technique was applied for scattered and bremsstrahlung photons. The resultant electrons from photon interactions are simulated by repeating the recorded electron tracks. The electron step length is stretched/shrunk proportionally based on the local density and stopping power ratios of the local material. Energy deposition in a voxel is proportional to the fraction of the equivalent step length in that voxel. To evaluate its accuracy, dose deposition in a 300mm*300mm*300mm water phantom is calculated, and compared to EGSnrc results. Results: Both PDD and OAR showed great agreements (within 0.5%) between our dose engine result and the EGSnrc result. It only takes less than 1 min for every simulation, being reduced up to ∼40 times compared to EGSnrc simulations. Conclusion: We have successfully developed a fast Monte Carlo dose engine for Gamma Knife.

PICA, Photon-Induced Medium-Range Nuclear Cascade Analysis by Monte-Carlo

International Nuclear Information System (INIS)

2001-01-01

1 - Description of program or function: PIC calculates the results of nuclear reactions caused by the collision of medium-energy photons with nuclei. The photon energy range in which the calculations are applicable is 30 4 are possible. The program PIC can accommodate incident monoenergetic photons as well as thin-target Bremsstrahlung spectra, thin-target Bremsstrahlung difference spectra and thick-target Bremsstrahlung spectra. For the last type of spectra the user must furnish the photon spectral data. PIC writes a history tape containing data on the properties of the particles (protons, neutrons, or pions) escaping from the nucleus. The data consists of the types of escaping particles and their energies and angles of emission. MECCAN utilizes the data on the PIC history tape to calculate cross sections such as the nonelastic cross section or the doubly differential cross section for energy-angle correlated distributions. EVAP then carries the nuclear reaction through an additional phase, that of evaporation, and calculates the energy spectra of particles (protons, neutrons, deuterons, tritons, 3 He, and alpha particles) 'boiled off' from the nucleus after the cascade has stopped, evaporation particle multiplicities, and evaporation residual nuclei (radio-chemical) cross sections. 2 - Method of solution: The interaction of high-energy photons with nuclei is described by using the intranuclear cascade and evaporation models. Monte Carlo methods are employed to provide a detailed description of each interaction. The initial interaction of the photon with the nucleus is obtained from the quasi-deuteron model of Levinger, that is, photon absorption by a neutron-proton pair moving within the nucleus or from one of the four pion-nucleon states formed in the photon-nucleon interaction. The effect of secondary nucleon-nucleus and/or pion-nucleus interactions following the photon absorption is accounted for by utilizing the intranuclear-cascade concept of high

Monte Carlo systems used for treatment planning and dose verification

Energy Technology Data Exchange (ETDEWEB)

Brualla, Lorenzo [Universitaetsklinikum Essen, NCTeam, Strahlenklinik, Essen (Germany); Rodriguez, Miguel [Centro Medico Paitilla, Balboa (Panama); Lallena, Antonio M. [Universidad de Granada, Departamento de Fisica Atomica, Molecular y Nuclear, Granada (Spain)

2017-04-15

General-purpose radiation transport Monte Carlo codes have been used for estimation of the absorbed dose distribution in external photon and electron beam radiotherapy patients since several decades. Results obtained with these codes are usually more accurate than those provided by treatment planning systems based on non-stochastic methods. Traditionally, absorbed dose computations based on general-purpose Monte Carlo codes have been used only for research, owing to the difficulties associated with setting up a simulation and the long computation time required. To take advantage of radiation transport Monte Carlo codes applied to routine clinical practice, researchers and private companies have developed treatment planning and dose verification systems that are partly or fully based on fast Monte Carlo algorithms. This review presents a comprehensive list of the currently existing Monte Carlo systems that can be used to calculate or verify an external photon and electron beam radiotherapy treatment plan. Particular attention is given to those systems that are distributed, either freely or commercially, and that do not require programming tasks from the end user. These systems are compared in terms of features and the simulation time required to compute a set of benchmark calculations. (orig.) [German] Seit mehreren Jahrzehnten werden allgemein anwendbare Monte-Carlo-Codes zur Simulation des Strahlungstransports benutzt, um die Verteilung der absorbierten Dosis in der perkutanen Strahlentherapie mit Photonen und Elektronen zu evaluieren. Die damit erzielten Ergebnisse sind meist akkurater als solche, die mit nichtstochastischen Methoden herkoemmlicher Bestrahlungsplanungssysteme erzielt werden koennen. Wegen des damit verbundenen Arbeitsaufwands und der langen Dauer der Berechnungen wurden Monte-Carlo-Simulationen von Dosisverteilungen in der konventionellen Strahlentherapie in der Vergangenheit im Wesentlichen in der Forschung eingesetzt. Im Bemuehen, Monte-Carlo

MCNP: a general Monte Carlo code for neutron and photon transport. Version 3A. Revision 2

International Nuclear Information System (INIS)

Briesmeister, J.F.

1986-09-01

This manual is a practical guide for the use of our general-purpose Monte Carlo code MCNP. The first chapter is a primer for the novice user. The second chapter describes the mathematics, data, physics, and Monte Carlo simulation found in MCNP. This discussion is not meant to be exhaustive - details of the particular techniques and of the Monte Carlo method itself will have to be found elsewhere. The third chapter shows the user how to prepare input for the code. The fourth chapter contains several examples, and the fifth chapter explains the output. The appendices show how to use MCNP on particular computer systems at the Los Alamos National Laboratory and also give details about some of the code internals that those who wish to modify the code may find useful. 57 refs

Electron transport in radiotherapy using local-to-global Monte Carlo

International Nuclear Information System (INIS)

Svatos, M.M.; Chandler, W.P.; Siantar, C.L.H.; Rathkopf, J.A.; Ballinger, C.T.

1994-09-01

Local-to-Global (L-G) Monte Carlo methods are a way to make three-dimensional electron transport both fast and accurate relative to other Monte Carlo methods. This is achieved by breaking the simulation into two stages: a local calculation done over small geometries having the size and shape of the ''steps'' to be taken through the mesh; and a global calculation which relies on a stepping code that samples the stored results of the local calculation. The increase in speed results from taking fewer steps in the global calculation than required by ordinary Monte Carlo codes and by speeding up the calculation per step. The potential for accuracy comes from the ability to use long runs of detailed codes to compile probability distribution functions (PDFs) in the local calculation. Specific examples of successful Local-to-Global algorithms are given

The ARGUS electron/photon calorimeter. Pt. 2

International Nuclear Information System (INIS)

Drescher, A.; Graf, H.J.; Graewe, B.; Hofmann, W.; Markees, A.; Matthiesen, U.; Spengler, J.; Wegener, D.

1983-03-01

A detailed investigation is described of the photon production and transport in lead scintillator shower counters with wave-length shifter readout built for the ARGUS detector. Experimental data and Monte Carlo calculations are in good agreement. The most prominent effects due to the light collection system are small nonlinearities in the relation between deposited energy and pulse height and an energy dependent decrease of the energy resolution of the counters. (orig.)

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

Transport methods: general. 2. Monte Carlo Particle Transport in Media with Exponentially Varying Time-Dependent Cross Sections

International Nuclear Information System (INIS)

Brown, Forrest B.; Martin, William R.

2001-01-01

We have investigated Monte Carlo schemes for analyzing particle transport through media with exponentially varying time-dependent cross sections. For such media, the cross sections are represented in the form Σ(t) = Σ 0 e -at (1) or equivalently as Σ(x) = Σ 0 e -bx (2) where b = av and v is the particle speed. For the following discussion, the parameters a and b may be either positive, for exponentially decreasing cross sections, or negative, for exponentially increasing cross sections. For most time-dependent Monte Carlo applications, the time and spatial variations of the cross-section data are handled by means of a stepwise procedure, holding the cross sections constant for each region over a small time interval Δt, performing the Monte Carlo random walk over the interval Δt, updating the cross sections, and then repeating for a series of time intervals. Continuously varying spatial- or time-dependent cross sections can be treated in a rigorous Monte Carlo fashion using delta-tracking, but inefficiencies may arise if the range of cross-section variation is large. In this paper, we present a new method for sampling collision distances directly for cross sections that vary exponentially in space or time. The method is exact and efficient and has direct application to Monte Carlo radiation transport methods. To verify that the probability density function (PDF) is correct and that the random-sampling procedure yields correct results, numerical experiments were performed using a one-dimensional Monte Carlo code. The physical problem consisted of a beam source impinging on a purely absorbing infinite slab, with a slab thickness of 1 cm and Σ 0 = 1 cm -1 . Monte Carlo calculations with 10 000 particles were run for a range of the exponential parameter b from -5 to +20 cm -1 . Two separate Monte Carlo calculations were run for each choice of b, a continuously varying case using the random-sampling procedures described earlier, and a 'conventional' case where the

Raman Monte Carlo simulation for light propagation for tissue with embedded objects

Science.gov (United States)

Periyasamy, Vijitha; Jaafar, Humaira Bte; Pramanik, Manojit

2018-02-01

Monte Carlo (MC) stimulation is one of the prominent simulation technique and is rapidly becoming the model of choice to study light-tissue interaction. Monte Carlo simulation for light transport in multi-layered tissue (MCML) is adapted and modelled with different geometry by integrating embedded objects of various shapes (i.e., sphere, cylinder, cuboid and ellipsoid) into the multi-layered structure. These geometries would be useful in providing a realistic tissue structure such as modelling for lymph nodes, tumors, blood vessels, head and other simulation medium. MC simulations were performed on various geometric medium. Simulation of MCML with embedded object (MCML-EO) was improvised for propagation of the photon in the defined medium with Raman scattering. The location of Raman photon generation is recorded. Simulations were experimented on a modelled breast tissue with tumor (spherical and ellipsoidal) and blood vessels (cylindrical). Results were presented in both A-line and B-line scans for embedded objects to determine spatial location where Raman photons were generated. Studies were done for different Raman probabilities.

Physical models, cross sections, and numerical approximations used in MCNP and GEANT4 Monte Carlo codes for photon and electron absorbed fraction calculation.

Science.gov (United States)

Yoriyaz, Hélio; Moralles, Maurício; Siqueira, Paulo de Tarso Dalledone; Guimarães, Carla da Costa; Cintra, Felipe Belonsi; dos Santos, Adimir

2009-11-01

Radiopharmaceutical applications in nuclear medicine require a detailed dosimetry estimate of the radiation energy delivered to the human tissues. Over the past years, several publications addressed the problem of internal dose estimate in volumes of several sizes considering photon and electron sources. Most of them used Monte Carlo radiation transport codes. Despite the widespread use of these codes due to the variety of resources and potentials they offered to carry out dose calculations, several aspects like physical models, cross sections, and numerical approximations used in the simulations still remain an object of study. Accurate dose estimate depends on the correct selection of a set of simulation options that should be carefully chosen. This article presents an analysis of several simulation options provided by two of the most used codes worldwide: MCNP and GEANT4. For this purpose, comparisons of absorbed fraction estimates obtained with different physical models, cross sections, and numerical approximations are presented for spheres of several sizes and composed as five different biological tissues. Considerable discrepancies have been found in some cases not only between the different codes but also between different cross sections and algorithms in the same code. Maximum differences found between the two codes are 5.0% and 10%, respectively, for photons and electrons. Even for simple problems as spheres and uniform radiation sources, the set of parameters chosen by any Monte Carlo code significantly affects the final results of a simulation, demonstrating the importance of the correct choice of parameters in the simulation.

Installation of Monte Carlo neutron and photon transport code system MCNP4

International Nuclear Information System (INIS)

Takano, Makoto; Sasaki, Mikio; Kaneko, Toshiyuki; Yamazaki, Takao.

1993-03-01

The continuous energy Monte Carlo code MCNP-4 including its graphic functions has been installed on the Sun-4 sparc-2 work station with minor corrections. In order to validate the installed MCNP-4 code, 25 sample problems have been executed on the work station and these results have been compared with the original ones. And, the most of the graphic functions have been demonstrated by using 3 sample problems. Further, additional 14 nuclides have been included to the continuous cross section library edited from JENDL-3. (author)

Neutron/photon/electron shielding study for a laser-fusion facility

International Nuclear Information System (INIS)

Thompson, W.L.

1977-01-01

A Monte Carlo shielding study encompassing neutron, photon, and electron transport has been conducted for the High Energy Gas Laser Facility at the Los Alamos Scientific Laboratory. This paper describes the application of the Monte Carlo technique and several variance reduction schemes to the study. The calculations involve a geometry which is complicated in all three dimensions, a very intense 14 MeV neutron source, skyshine and deep penetrations. The facility design with 1.83 m concrete walls and a 1.52 m concrete roof is based on these calculations

The use of Monte-Carlo codes for treatment planning in external-beam radiotherapy

International Nuclear Information System (INIS)

Alan, E.; Nahum, PhD.

2003-01-01

Monte Carlo simulation of radiation transport is a very powerful technique. There are basically no exact solutions to the Boltzmann transport equation. Even, the 'straightforward' situation (in radiotherapy) of an electron beam depth-dose distribution in water proves to be too difficult for analytical methods without making gross approximations such as ignoring energy-loss straggling, large-angle single scattering and Bremsstrahlung production. monte Carlo is essential when radiation is transport from one medium into another. As the particle (be it a neutron, photon, electron, proton) crosses the boundary then a new set of interaction cross-sections is simply read in and the simulation continues as though the new medium were infinite until the next boundary is encountered. Radiotherapy involves directing a beam of megavoltage x rays or electrons (occasionally protons) at a very complex object, the human body. Monte Carlo simulation has proved in valuable at many stages of the process of accurately determining the distribution of absorbed dose in the patient. Some of these applications will be reviewed here. (Rogers and al 1990; Andreo 1991; Mackie 1990). (N.C.)

Monte Carlo simulation for the transport beamline

Energy Technology Data Exchange (ETDEWEB)

Romano, F.; Cuttone, G.; Jia, S. B.; Varisano, A. [INFN, Laboratori Nazionali del Sud, Via Santa Sofia 62, Catania (Italy); Attili, A.; Marchetto, F.; Russo, G. [INFN, Sezione di Torino, Via P.Giuria, 1 10125 Torino (Italy); Cirrone, G. A. P.; Schillaci, F.; Scuderi, V. [INFN, Laboratori Nazionali del Sud, Via Santa Sofia 62, Catania, Italy and Institute of Physics Czech Academy of Science, ELI-Beamlines project, Na Slovance 2, Prague (Czech Republic); Carpinelli, M. [INFN Sezione di Cagliari, c/o Dipartimento di Fisica, Università di Cagliari, Cagliari (Italy); Tramontana, A. [INFN, Laboratori Nazionali del Sud, Via Santa Sofia 62, Catania, Italy and Università di Catania, Dipartimento di Fisica e Astronomia, Via S. Sofia 64, Catania (Italy)

2013-07-26

In the framework of the ELIMED project, Monte Carlo (MC) simulations are widely used to study the physical transport of charged particles generated by laser-target interactions and to preliminarily evaluate fluence and dose distributions. An energy selection system and the experimental setup for the TARANIS laser facility in Belfast (UK) have been already simulated with the GEANT4 (GEometry ANd Tracking) MC toolkit. Preliminary results are reported here. Future developments are planned to implement a MC based 3D treatment planning in order to optimize shots number and dose delivery.

Monte Carlo simulation for the transport beamline

International Nuclear Information System (INIS)

Romano, F.; Cuttone, G.; Jia, S. B.; Varisano, A.; Attili, A.; Marchetto, F.; Russo, G.; Cirrone, G. A. P.; Schillaci, F.; Scuderi, V.; Carpinelli, M.; Tramontana, A.

2013-01-01

In the framework of the ELIMED project, Monte Carlo (MC) simulations are widely used to study the physical transport of charged particles generated by laser-target interactions and to preliminarily evaluate fluence and dose distributions. An energy selection system and the experimental setup for the TARANIS laser facility in Belfast (UK) have been already simulated with the GEANT4 (GEometry ANd Tracking) MC toolkit. Preliminary results are reported here. Future developments are planned to implement a MC based 3D treatment planning in order to optimize shots number and dose delivery

Photon dose estimation from ultraintense laser–solid interactions and shielding calculation with Monte Carlo simulation

International Nuclear Information System (INIS)

Yang, Bo; Qiu, Rui; Li, JunLi; Lu, Wei; Wu, Zhen; Li, Chunyan

2017-01-01

When a strong laser beam irradiates a solid target, a hot plasma is produced and high-energy electrons are usually generated (the so-called “hot electrons”). These energetic electrons subsequently generate hard X-rays in the solid target through the Bremsstrahlung process. To date, only limited studies have been conducted on this laser-induced radiological protection issue. In this study, extensive literature reviews on the physics and properties of hot electrons have been conducted. On the basis of these information, the photon dose generated by the interaction between hot electrons and a solid target was simulated with the Monte Carlo code FLUKA. With some reasonable assumptions, the calculated dose can be regarded as the upper boundary of the experimental results over the laser intensity ranging from 10 19 to 10 21 W/cm 2 . Furthermore, an equation to estimate the photon dose generated from ultraintense laser–solid interactions based on the normalized laser intensity is derived. The shielding effects of common materials including concrete and lead were also studied for the laser-driven X-ray source. The dose transmission curves and tenth-value layers (TVLs) in concrete and lead were calculated through Monte Carlo simulations. These results could be used to perform a preliminary and fast radiation safety assessment for the X-rays generated from ultraintense laser–solid interactions. - Highlights: • The laser–driven X-ray ionizing radiation source was analyzed in this study. • An equation to estimate the photon dose based on the laser intensity is given. • The shielding effects of concrete and lead were studied for this new X-ray source. • The aim of this study is to analyze and mitigate the laser–driven X-ray hazard.

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)

Monte Carlo treatment planning with modulated electron radiotherapy: framework development and application

Science.gov (United States)

Alexander, Andrew William

Within the field of medical physics, Monte Carlo radiation transport simulations are considered to be the most accurate method for the determination of dose distributions in patients. The McGill Monte Carlo treatment planning system (MMCTP), provides a flexible software environment to integrate Monte Carlo simulations with current and new treatment modalities. A developing treatment modality called energy and intensity modulated electron radiotherapy (MERT) is a promising modality, which has the fundamental capabilities to enhance the dosimetry of superficial targets. An objective of this work is to advance the research and development of MERT with the end goal of clinical use. To this end, we present the MMCTP system with an integrated toolkit for MERT planning and delivery of MERT fields. Delivery is achieved using an automated "few leaf electron collimator" (FLEC) and a controller. Aside from the MERT planning toolkit, the MMCTP system required numerous add-ons to perform the complex task of large-scale autonomous Monte Carlo simulations. The first was a DICOM import filter, followed by the implementation of DOSXYZnrc as a dose calculation engine and by logic methods for submitting and updating the status of Monte Carlo simulations. Within this work we validated the MMCTP system with a head and neck Monte Carlo recalculation study performed by a medical dosimetrist. The impact of MMCTP lies in the fact that it allows for systematic and platform independent large-scale Monte Carlo dose calculations for different treatment sites and treatment modalities. In addition to the MERT planning tools, various optimization algorithms were created external to MMCTP. The algorithms produced MERT treatment plans based on dose volume constraints that employ Monte Carlo pre-generated patient-specific kernels. The Monte Carlo kernels are generated from patient-specific Monte Carlo dose distributions within MMCTP. The structure of the MERT planning toolkit software and

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.

Microwave transport in EBT distribution manifolds using Monte Carlo ray-tracing techniques

International Nuclear Information System (INIS)

Lillie, R.A.; White, T.L.; Gabriel, T.A.; Alsmiller, R.G. Jr.

1983-01-01

Ray tracing Monte Carlo calculations have been carried out using an existing Monte Carlo radiation transport code to obtain estimates of the microsave power exiting the torus coupling links in EPT microwave manifolds. The microwave power loss and polarization at surface reflections were accounted for by treating the microwaves as plane waves reflecting off plane surfaces. Agreement on the order of 10% was obtained between the measured and calculated output power distribution for an existing EBT-S toroidal manifold. A cost effective iterative procedure utilizing the Monte Carlo history data was implemented to predict design changes which could produce increased manifold efficiency and improved output power uniformity

Transport and hydrodynamic calculations of direct photons at FAIR

International Nuclear Information System (INIS)

Baeuchle, Bjorn; Bleicher, Marcus

2011-01-01

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

Verification of the VEF photon beam model for dose calculations by the voxel-Monte-Carlo-algorithm

International Nuclear Information System (INIS)

Kriesen, S.; Fippel, M.

2005-01-01

The VEF linac head model (VEF, virtual energy fluence) was developed at the University of Tuebingen to determine the primary fluence for calculations of dose distributions in patients by the Voxel-Monte-Carlo-Algorithm (XVMC). This analytical model can be fitted to any therapy accelerator head by measuring only a few basic dose data; therefore, time-consuming Monte-Carlo simulations of the linac head become unnecessary. The aim of the present study was the verification of the VEF model by means of water-phantom measurements, as well as the comparison of this system with a common analytical linac head model of a commercial planning system (TMS, formerly HELAX or MDS Nordion, respectively). The results show that both the VEF and the TMS models can very well simulate the primary fluence. However, the VEF model proved superior in the simulations of scattered radiation and in the calculations of strongly irregular MLC fields. Thus, an accurate and clinically practicable tool for the determination of the primary fluence for Monte-Carlo-Simulations with photons was established, especially for the use in IMRT planning. (orig.)

[Verification of the VEF photon beam model for dose calculations by the Voxel-Monte-Carlo-Algorithm].

Science.gov (United States)

Kriesen, Stephan; Fippel, Matthias

2005-01-01

The VEF linac head model (VEF, virtual energy fluence) was developed at the University of Tübingen to determine the primary fluence for calculations of dose distributions in patients by the Voxel-Monte-Carlo-Algorithm (XVMC). This analytical model can be fitted to any therapy accelerator head by measuring only a few basic dose data; therefore, time-consuming Monte-Carlo simulations of the linac head become unnecessary. The aim of the present study was the verification of the VEF model by means of water-phantom measurements, as well as the comparison of this system with a common analytical linac head model of a commercial planning system (TMS, formerly HELAX or MDS Nordion, respectively). The results show that both the VEF and the TMS models can very well simulate the primary fluence. However, the VEF model proved superior in the simulations of scattered radiation and in the calculations of strongly irregular MLC fields. Thus, an accurate and clinically practicable tool for the determination of the primary fluence for Monte-Carlo-Simulations with photons was established, especially for the use in IMRT planning.

A Monte Carlo multiple source model applied to radiosurgery narrow photon beams

International Nuclear Information System (INIS)

Chaves, A.; Lopes, M.C.; Alves, C.C.; Oliveira, C.; Peralta, L.; Rodrigues, P.; Trindade, A.

2004-01-01

Monte Carlo (MC) methods are nowadays often used in the field of radiotherapy. Through successive steps, radiation fields are simulated, producing source Phase Space Data (PSD) that enable a dose calculation with good accuracy. Narrow photon beams used in radiosurgery can also be simulated by MC codes. However, the poor efficiency in simulating these narrow photon beams produces PSD whose quality prevents calculating dose with the required accuracy. To overcome this difficulty, a multiple source model was developed that enhances the quality of the reconstructed PSD, reducing also the time and storage capacities. This multiple source model was based on the full MC simulation, performed with the MC code MCNP4C, of the Siemens Mevatron KD2 (6 MV mode) linear accelerator head and additional collimators. The full simulation allowed the characterization of the particles coming from the accelerator head and from the additional collimators that shape the narrow photon beams used in radiosurgery treatments. Eight relevant photon virtual sources were identified from the full characterization analysis. Spatial and energy distributions were stored in histograms for the virtual sources representing the accelerator head components and the additional collimators. The photon directions were calculated for virtual sources representing the accelerator head components whereas, for the virtual sources representing the additional collimators, they were recorded into histograms. All these histograms were included in the MC code, DPM code and using a sampling procedure that reconstructed the PSDs, dose distributions were calculated in a water phantom divided in 20000 voxels of 1x1x5 mm 3 . The model accurately calculates dose distributions in the water phantom for all the additional collimators; for depth dose curves, associated errors at 2σ were lower than 2.5% until a depth of 202.5 mm for all the additional collimators and for profiles at various depths, deviations between measured

Evaluation of a 50-MV photon therapy beam from a racetrack microtron using MCNP4B Monte Carlo code

International Nuclear Information System (INIS)

Gudowska, I.; Svensson, R.

2001-01-01

High energy photon therapy beam from the 50 MV racetrack microtron has been evaluated using the Monte Carlo code MCNP4B. The spatial and energy distribution of photons, radial and depth dose distributions in the phantom are calculated for the stationary and scanned photon beams from different targets. The calculated dose distributions are compared to the experimental data using a silicon diode detector. Measured and calculated depth-dose distributions are in fairly good agreement, within 2-3% for the positions in the range 2-30 cm in the phantom, whereas the larger discrepancies up to 10% are observed in the dose build-up region. For the stationary beams the differences in the calculated and measured radial dose distributions are about 2-10%. (orig.)

Hybrid transport and diffusion modeling using electron thermal transport Monte Carlo SNB in DRACO

Science.gov (United States)

Chenhall, Jeffrey; Moses, Gregory

2017-10-01

The iSNB (implicit Schurtz Nicolai Busquet) multigroup diffusion electron thermal transport method is adapted into an Electron Thermal Transport Monte Carlo (ETTMC) transport method to better model angular and long mean free path non-local effects. Previously, the ETTMC model had been implemented in the 2D DRACO multiphysics code and found to produce consistent results with the iSNB method. Current work is focused on a hybridization of the computationally slower but higher fidelity ETTMC transport method with the computationally faster iSNB diffusion method in order to maximize computational efficiency. Furthermore, effects on the energy distribution of the heat flux divergence are studied. Work to date on the hybrid method will be presented. This work was supported by Sandia National Laboratories and the Univ. of Rochester Laboratory for Laser Energetics.

ZZ ENDLIB, Coupled Electron and Photon Transport Library in ENDL Format

International Nuclear Information System (INIS)

2002-01-01

Description of program or function: The LLNL Evaluated Nuclear Data Library has existed since 1958 in a succession of forms and formats. The present form is as a machine-independent character file format and contains data for the evaluated atomic relaxation data library (EADL), the evaluated photon interaction data library (EPDL), and the evaluated electron interaction data library (EEDL). The purpose of these libraries is to furnish data for coupled electron-photon transport calculations. In order to perform coupled photon-electron transport calculations, all three libraries are required. The UCRL-ID-117796 report included in the documentation for this package provides information on the contents and formats for all three libraries, which are included in this package. All of these libraries span atomic numbers, Z, from 1 to 100. Additionally the electron and photon interaction libraries cover the incident particle energy range from 10 eV to 100 GeV

3D Monte Carlo model of optical transport in laser-irradiated cutaneous vascular malformations

Science.gov (United States)

Majaron, Boris; Milanič, Matija; Jia, Wangcun; Nelson, J. S.

2010-11-01

We have developed a three-dimensional Monte Carlo (MC) model of optical transport in skin and applied it to analysis of port wine stain treatment with sequential laser irradiation and intermittent cryogen spray cooling. Our MC model extends the approaches of the popular multi-layer model by Wang et al.1 to three dimensions, thus allowing treatment of skin inclusions with more complex geometries and arbitrary irradiation patterns. To overcome the obvious drawbacks of either "escape" or "mirror" boundary conditions at the lateral boundaries of the finely discretized volume of interest (VOI), photons exiting the VOI are propagated in laterally infinite tissue layers with appropriate optical properties, until they loose all their energy, escape into the air, or return to the VOI, but the energy deposition outside of the VOI is not computed and recorded. After discussing the selection of tissue parameters, we apply the model to analysis of blood photocoagulation and collateral thermal damage in treatment of port wine stain (PWS) lesions with sequential laser irradiation and intermittent cryogen spray cooling.

KAMCCO, a reactor physics Monte Carlo neutron transport code

International Nuclear Information System (INIS)

Arnecke, G.; Borgwaldt, H.; Brandl, V.; Lalovic, M.

1976-06-01

KAMCCO is a 3-dimensional reactor Monte Carlo code for fast neutron physics problems. Two options are available for the solution of 1) the inhomogeneous time-dependent neutron transport equation (census time scheme), and 2) the homogeneous static neutron transport equation (generation cycle scheme). The user defines the desired output, e.g. estimates of reaction rates or neutron flux integrated over specified volumes in phase space and time intervals. Such primary quantities can be arbitrarily combined, also ratios of these quantities can be estimated with their errors. The Monte Carlo techniques are mostly analogue (exceptions: Importance sampling for collision processes, ELP/MELP, Russian roulette and splitting). Estimates are obtained from the collision and track length estimators. Elastic scattering takes into account first order anisotropy in the center of mass system. Inelastic scattering is processed via the evaporation model or via the excitation of discrete levels. For the calculation of cross sections, the energy is treated as a continuous variable. They are computed by a) linear interpolation, b) from optionally Doppler broadened single level Breit-Wigner resonances or c) from probability tables (in the region of statistically distributed resonances). (orig.) [de

Modification to the Monte Carlo N-Particle (MCNP) Visual Editor (MCNPVised) to Read in Computer Aided Design (CAD) Files

International Nuclear Information System (INIS)

Randolph Schwarz; Leland L. Carter; Alysia Schwarz

2005-01-01

Monte Carlo N-Particle Transport Code (MCNP) is the code of choice for doing complex neutron/photon/electron transport calculations for the nuclear industry and research institutions. The Visual Editor for Monte Carlo N-Particle is internationally recognized as the best code for visually creating and graphically displaying input files for MCNP. The work performed in this grant was used to enhance the capabilities of the MCNP Visual Editor to allow it to read in both 2D and 3D Computer Aided Design (CAD) files, allowing the user to electronically generate a valid MCNP input geometry

A deterministic electron, photon, proton and heavy ion transport suite for the study of the Jovian moon Europa

International Nuclear Information System (INIS)

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

2011-01-01

A Langley research center (LaRC) developed deterministic suite of radiation transport codes describing the propagation of electron, photon, proton and heavy ion in condensed media is used to simulate the exposure from the spectral distribution of the aforementioned particles in the Jovian radiation environment. Based on the measurements by the Galileo probe (1995-2003) heavy ion counter (HIC), the choice of trapped heavy ions is limited to carbon, oxygen and sulfur (COS). The deterministic particle transport suite consists of a coupled electron photon algorithm (CEPTRN) and a coupled light heavy ion algorithm (HZETRN). The primary purpose for the development of the transport suite is to provide a means to the spacecraft design community to rapidly perform numerous repetitive calculations essential for electron, photon, proton and heavy ion exposure assessment in a complex space structure. In this paper, the reference 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 and proton 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 JPL provided Europa mission fluence spectrum, is used to produce the corresponding depth dose curve in silicon behind a default aluminum shield of 100 mils (∼0.7 g/cm 2 ). The transport suite can also accept a geometry describing ray traced thickness file from a computer aided design (CAD) package and calculate the total ionizing dose (TID) at a specific target point within the interior of the vehicle. In that regard, using a low fidelity CAD model of the Galileo probe generated by the authors, the transport suite was verified versus Monte Carlo (MC) simulation for orbits JOI-J35 of the Galileo probe

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

Automatic modeling for the monte carlo transport TRIPOLI code

International Nuclear Information System (INIS)

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

2010-01-01

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

MC 93 - Proceedings of the International Conference on Monte Carlo Simulation in High Energy and Nuclear Physics

Science.gov (United States)

Dragovitsch, Peter; Linn, Stephan L.; Burbank, Mimi

1994-01-01

The Table of Contents for the book is as follows: * Preface * Heavy Fragment Production for Hadronic Cascade Codes * Monte Carlo Simulations of Space Radiation Environments * Merging Parton Showers with Higher Order QCD Monte Carlos * An Order-αs Two-Photon Background Study for the Intermediate Mass Higgs Boson * GEANT Simulation of Hall C Detector at CEBAF * Monte Carlo Simulations in Radioecology: Chernobyl Experience * UNIMOD2: Monte Carlo Code for Simulation of High Energy Physics Experiments; Some Special Features * Geometrical Efficiency Analysis for the Gamma-Neutron and Gamma-Proton Reactions * GISMO: An Object-Oriented Approach to Particle Transport and Detector Modeling * Role of MPP Granularity in Optimizing Monte Carlo Programming * Status and Future Trends of the GEANT System * The Binary Sectioning Geometry for Monte Carlo Detector Simulation * A Combined HETC-FLUKA Intranuclear Cascade Event Generator * The HARP Nucleon Polarimeter * Simulation and Data Analysis Software for CLAS * TRAP -- An Optical Ray Tracing Program * Solutions of Inverse and Optimization Problems in High Energy and Nuclear Physics Using Inverse Monte Carlo * FLUKA: Hadronic Benchmarks and Applications * Electron-Photon Transport: Always so Good as We Think? Experience with FLUKA * Simulation of Nuclear Effects in High Energy Hadron-Nucleus Collisions * Monte Carlo Simulations of Medium Energy Detectors at COSY Jülich * Complex-Valued Monte Carlo Method and Path Integrals in the Quantum Theory of Localization in Disordered Systems of Scatterers * Radiation Levels at the SSCL Experimental Halls as Obtained Using the CLOR89 Code System * Overview of Matrix Element Methods in Event Generation * Fast Electromagnetic Showers * GEANT Simulation of the RMC Detector at TRIUMF and Neutrino Beams for KAON * Event Display for the CLAS Detector * Monte Carlo Simulation of High Energy Electrons in Toroidal Geometry * GEANT 3.14 vs. EGS4: A Comparison Using the DØ Uranium/Liquid Argon

MCNP/X TRANSPORT IN THE TABULAR REGIME

Energy Technology Data Exchange (ETDEWEB)

HUGHES, H. GRADY [Los Alamos National Laboratory

2007-01-08

The authors review the transport capabilities of the MCNP and MCNPX Monte Carlo codes in the energy regimes in which tabular transport data are available. Giving special attention to neutron tables, they emphasize the measures taken to improve the treatment of a variety of difficult aspects of the transport problem, including unresolved resonances, thermal issues, and the availability of suitable cross sections sets. They also briefly touch on the current situation in regard to photon, electron, and proton transport tables.

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

CERN Document Server

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

2001-01-01

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

Monte Carlo simulation of the response of a pixellated 3D photo-detector in silicon

CERN Document Server

Dubaric, E; Froejdh, C; Norlin, B

2002-01-01

The charge transport and X-ray photon absorption in three-dimensional (3D) X-ray pixel detectors have been studied using numerical simulations. The charge transport has been modelled using the drift-diffusion simulator MEDICI, while photon absorption has been studied using MCNP. The response of the entire pixel detector system in terms of charge sharing, line spread function and modulation transfer function, has been simulated using a system level Monte Carlo simulation approach. A major part of the study is devoted to the effect of charge sharing on the energy resolution in 3D-pixel detectors. The 3D configuration was found to suppress charge sharing much better than conventional planar detectors.

Response matrix Monte Carlo based on a general geometry local calculation for electron transport

International Nuclear Information System (INIS)

Ballinger, C.T.; Rathkopf, J.A.; Martin, W.R.

1991-01-01

A Response Matrix Monte Carlo (RMMC) method has been developed for solving electron transport problems. This method was born of the need to have a reliable, computationally efficient transport method for low energy electrons (below a few hundred keV) in all materials. Today, condensed history methods are used which reduce the computation time by modeling the combined effect of many collisions but fail at low energy because of the assumptions required to characterize the electron scattering. Analog Monte Carlo simulations are prohibitively expensive since electrons undergo coulombic scattering with little state change after a collision. The RMMC method attempts to combine the accuracy of an analog Monte Carlo simulation with the speed of the condensed history methods. Like condensed history, the RMMC method uses probability distributions functions (PDFs) to describe the energy and direction of the electron after several collisions. However, unlike the condensed history method the PDFs are based on an analog Monte Carlo simulation over a small region. Condensed history theories require assumptions about the electron scattering to derive the PDFs for direction and energy. Thus the RMMC method samples from PDFs which more accurately represent the electron random walk. Results show good agreement between the RMMC method and analog Monte Carlo. 13 refs., 8 figs

The OpenMC Monte Carlo particle transport code

International Nuclear Information System (INIS)

Romano, Paul K.; Forget, Benoit

2013-01-01

Highlights: ► An open source Monte Carlo particle transport code, OpenMC, has been developed. ► Solid geometry and continuous-energy physics allow high-fidelity simulations. ► Development has focused on high performance and modern I/O techniques. ► OpenMC is capable of scaling up to hundreds of thousands of processors. ► Results on a variety of benchmark problems agree with MCNP5. -- Abstract: A new Monte Carlo code called OpenMC is currently under development at the Massachusetts Institute of Technology as a tool for simulation on high-performance computing platforms. Given that many legacy codes do not scale well on existing and future parallel computer architectures, OpenMC has been developed from scratch with a focus on high performance scalable algorithms as well as modern software design practices. The present work describes the methods used in the OpenMC code and demonstrates the performance and accuracy of the code on a variety of problems.

Exponentially-convergent Monte Carlo for the 1-D transport equation

International Nuclear Information System (INIS)

Peterson, J. R.; Morel, J. E.; Ragusa, J. C.

2013-01-01

We define a new exponentially-convergent Monte Carlo method for solving the one-speed 1-D slab-geometry transport equation. This method is based upon the use of a linear discontinuous finite-element trial space in space and direction to represent the transport solution. A space-direction h-adaptive algorithm is employed to restore exponential convergence after stagnation occurs due to inadequate trial-space resolution. This methods uses jumps in the solution at cell interfaces as an error indicator. Computational results are presented demonstrating the efficacy of the new approach. (authors)

Monte Carlo simulation of nonlinear reactive contaminant transport in unsaturated porous media

International Nuclear Information System (INIS)

Giacobbo, F.; Patelli, E.

2007-01-01

In the current proposed solutions of radioactive waste repositories, the protective function against the radionuclide water-driven transport back to the biosphere is to be provided by an integrated system of engineered and natural geologic barriers. The occurrence of several nonlinear interactions during the radionuclide migration process may render burdensome the classical analytical-numerical approaches. Moreover, the heterogeneity of the barriers' media forces approximations to the classical analytical-numerical models, thus reducing their fidelity to reality. In an attempt to overcome these difficulties, in the present paper we adopt a Monte Carlo simulation approach, previously developed on the basis of the Kolmogorov-Dmitriev theory of branching stochastic processes. The approach is here extended for describing transport through unsaturated porous media under transient flow conditions and in presence of nonlinear interchange phenomena between the liquid and solid phases. This generalization entails the determination of the functional dependence of the parameters of the proposed transport model from the water content and from the contaminant concentration, which change in space and time during the water infiltration process. The corresponding Monte Carlo simulation approach is verified with respect to a case of nonreactive transport under transient unsaturated flow and to a case of nonlinear reactive transport under stationary saturated flow. Numerical applications regarding linear and nonlinear reactive transport under transient unsaturated flow are reported

Verification of Monte Carlo transport codes by activation experiments

Energy Technology Data Exchange (ETDEWEB)

Chetvertkova, Vera

2012-12-18

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

Validation of cross sections for Monte Carlo simulation of the photoelectric effect

CERN Document Server

Han, Min Cheol; Pia, Maria Grazia; Basaglia, Tullio; Batic, Matej; Hoff, Gabriela; Kim, Chan Hyeong; Saracco, Paolo

2016-01-01

Several total and partial photoionization cross section calculations, based on both theoretical and empirical approaches, are quantitatively evaluated with statistical analyses using a large collection of experimental data retrieved from the literature to identify the state of the art for modeling the photoelectric effect in Monte Carlo particle transport. Some of the examined cross section models are available in general purpose Monte Carlo systems, while others have been implemented and subjected to validation tests for the first time to estimate whether they could improve the accuracy of particle transport codes. The validation process identifies Scofield's 1973 non-relativistic calculations, tabulated in the Evaluated Photon Data Library(EPDL), as the one best reproducing experimental measurements of total cross sections. Specialized total cross section models, some of which derive from more recent calculations, do not provide significant improvements. Scofield's non-relativistic calculations are not surp...

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

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

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

International Nuclear Information System (INIS)

Densmore, Jeffery D.; Thompson, Kelly G.; Urbatsch, Todd J.

2012-01-01

Discrete Diffusion Monte Carlo (DDMC) is a technique for increasing the efficiency of Implicit Monte Carlo radiative-transfer simulations in optically thick media. In DDMC, particles take discrete steps between spatial cells according to a discretized diffusion equation. Each discrete step replaces many smaller Monte Carlo steps, thus improving the efficiency of the simulation. In this paper, we present an extension of DDMC for frequency-dependent radiative transfer. We base our new DDMC method on a frequency-integrated diffusion equation for frequencies below a specified threshold, as optical thickness is typically a decreasing function of frequency. Above this threshold we employ standard Monte Carlo, which results in a hybrid transport-diffusion scheme. With a set of frequency-dependent test problems, we confirm the accuracy and increased efficiency of our new DDMC method.

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)

PENGEOM-A general-purpose geometry package for Monte Carlo simulation of radiation transport in material systems defined by quadric surfaces

Science.gov (United States)

Almansa, Julio; Salvat-Pujol, Francesc; Díaz-Londoño, Gloria; Carnicer, Artur; Lallena, Antonio M.; Salvat, Francesc

2016-02-01

The Fortran subroutine package PENGEOM provides a complete set of tools to handle quadric geometries in Monte Carlo simulations of radiation transport. The material structure where radiation propagates is assumed to consist of homogeneous bodies limited by quadric surfaces. The PENGEOM subroutines (a subset of the PENELOPE code) track particles through the material structure, independently of the details of the physics models adopted to describe the interactions. Although these subroutines are designed for detailed simulations of photon and electron transport, where all individual interactions are simulated sequentially, they can also be used in mixed (class II) schemes for simulating the transport of high-energy charged particles, where the effect of soft interactions is described by the random-hinge method. The definition of the geometry and the details of the tracking algorithm are tailored to optimize simulation speed. The use of fuzzy quadric surfaces minimizes the impact of round-off errors. The provided software includes a Java graphical user interface for editing and debugging the geometry definition file and for visualizing the material structure. Images of the structure are generated by using the tracking subroutines and, hence, they describe the geometry actually passed to the simulation code.

The FERMI-Elettra FEL Photon Transport System

International Nuclear Information System (INIS)

Zangrando, M.; Cudin, I.; Fava, C.; Godnig, R.; Kiskinova, M.; Masciovecchio, C.; Parmigiani, F.; Rumiz, L.; Svetina, C.; Turchet, A.; Cocco, D.

2010-01-01

The FERMI-Elettra free electron laser (FEL) user facility is under construction at Sincrotrone Trieste (Italy), and it will be operative in late 2010. It is based on a seeded scheme providing an almost perfect transform-limited and fully spatially coherent photon beam. FERMI-Elettra will cover the wavelength range 100 to 3 nm with the fundamental harmonics, and down to 1 nm with higher harmonics. We present the layout of the photon beam transport system that includes: the first common part providing on-line and shot-to-shot beam diagnostics, called PADReS (Photon Analysis Delivery and Reduction System), and 3 independent beamlines feeding the experimental stations. Particular emphasis is given to the solutions adopted to preserve the wavefront, and to avoid damage on the different optical elements. Peculiar FEL devices, not common in the Synchrotron Radiation facilities, are described in more detail, e.g. the online photon energy spectrometer measuring shot-by-shot the spectrum of the emitted radiation, the beam splitting and delay line system dedicated to cross/auto correlation and pump-probe experiments, and the wavefront preserving active optics adapting the shape and size of the focused spot to meet the needs of the different experiments.

A computationally efficient moment-preserving Monte Carlo electron transport method with implementation in Geant4

Energy Technology Data Exchange (ETDEWEB)

Dixon, D.A., E-mail: ddixon@lanl.gov [Los Alamos National Laboratory, P.O. Box 1663, MS P365, Los Alamos, NM 87545 (United States); Prinja, A.K., E-mail: prinja@unm.edu [Department of Nuclear Engineering, MSC01 1120, 1 University of New Mexico, Albuquerque, NM 87131-0001 (United States); Franke, B.C., E-mail: bcfrank@sandia.gov [Sandia National Laboratories, Albuquerque, NM 87123 (United States)

2015-09-15

This paper presents the theoretical development and numerical demonstration of a moment-preserving Monte Carlo electron transport method. Foremost, a full implementation of the moment-preserving (MP) method within the Geant4 particle simulation toolkit is demonstrated. Beyond implementation details, it is shown that the MP method is a viable alternative to the condensed history (CH) method for inclusion in current and future generation transport codes through demonstration of the key features of the method including: systematically controllable accuracy, computational efficiency, mathematical robustness, and versatility. A wide variety of results common to electron transport are presented illustrating the key features of the MP method. In particular, it is possible to achieve accuracy that is statistically indistinguishable from analog Monte Carlo, while remaining up to three orders of magnitude more efficient than analog Monte Carlo simulations. Finally, it is shown that the MP method can be generalized to any applicable analog scattering DCS model by extending previous work on the MP method beyond analytical DCSs to the partial-wave (PW) elastic tabulated DCS data.

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

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

International Nuclear Information System (INIS)

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

2003-01-01

This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k eff (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)

The difference of scoring dose to water or tissues in Monte Carlo dose calculations for low energy brachytherapy photon sources.

Science.gov (United States)

Landry, Guillaume; Reniers, Brigitte; Pignol, Jean-Philippe; Beaulieu, Luc; Verhaegen, Frank

2011-03-01

The goal of this work is to compare D(m,m) (radiation transported in medium; dose scored in medium) and D(w,m) (radiation transported in medium; dose scored in water) obtained from Monte Carlo (MC) simulations for a subset of human tissues of interest in low energy photon brachytherapy. Using low dose rate seeds and an electronic brachytherapy source (EBS), the authors quantify the large cavity theory conversion factors required. The authors also assess whether ap plying large cavity theory utilizing the sources' initial photon spectra and average photon energy induces errors related to spatial spectral variations. First, ideal spherical geometries were investigated, followed by clinical brachytherapy LDR seed implants for breast and prostate cancer patients. Two types of dose calculations are performed with the GEANT4 MC code. (1) For several human tissues, dose profiles are obtained in spherical geometries centered on four types of low energy brachytherapy sources: 125I, 103Pd, and 131Cs seeds, as well as an EBS operating at 50 kV. Ratios of D(w,m) over D(m,m) are evaluated in the 0-6 cm range. In addition to mean tissue composition, compositions corresponding to one standard deviation from the mean are also studied. (2) Four clinical breast (using 103Pd) and prostate (using 125I) brachytherapy seed implants are considered. MC dose calculations are performed based on postimplant CT scans using prostate and breast tissue compositions. PTV D90 values are compared for D(w,m) and D(m,m). (1) Differences (D(w,m)/D(m,m)-1) of -3% to 70% are observed for the investigated tissues. For a given tissue, D(w,m)/D(m,m) is similar for all sources within 4% and does not vary more than 2% with distance due to very moderate spectral shifts. Variations of tissue composition about the assumed mean composition influence the conversion factors up to 38%. (2) The ratio of D90(w,m) over D90(m,m) for clinical implants matches D(w,m)/D(m,m) at 1 cm from the single point sources, Given

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-07-01

This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k{sub eff} (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)

Observation of Jet Photoproduction and Comparison to Monte Carlo Simulation

Energy Technology Data Exchange (ETDEWEB)

Lincoln, Donald W. [Rice Univ., Houston, TX (United States)

1994-01-01

The photon is the carrier of the electromagnetic force. However in addition to its well known nature, the theories of QCD and quantum mechanics would indicate that the photon can also for brief periods of time split into a $q\\bar{q}$ pair (an extended photon.) How these constituents share energy and momentum is an interesting question and such a measurement was investigated by scattering photons off protons. The post collision kinematics should reveal pre-collision information. Unfortunately, when these constituents exit the collision point, they undergo subsequent interactions (gluon radiation, fragmentation, etc.) which scramble their kinematics. An algorithm was explored which was shown via Monte Carlo techniques to partially disentangle these post collision interactions and reveal the collision kinematics. The presence or absence of large transverse momenta internal ($k_\\perp$) to the photon has a significant impact on the ability to reconstruct the kinematics of the leading order calculation hard scatter system. Reconstruction of the next to leading order high $E_\\perp$ partons is more straightforward. Since the photon exhibits this unusual behavior only part of the time, many of the collisions recorded will be with a non-extended (or direct) photon. Unless a method for culling only the extended photons out can be invented, this contamination of direct photons must be accounted for. No such culling method is currently known, and so any measurement will necessarily contain both photon types. Theoretical predictions using Monte Carlo methods are compared with the data and are found to reproduce many experimentally measured distributions quite well. Overall the LUND Monte Carlo reproduces the data better than the HERWIG Monte Carlo. As expected at low jet $E_\\perp$, the data set seems to be dominated by extended photons, with the mix becoming nearly equal at jet $E_\\perp > 4$ GeV. The existence of a large photon $k_\\perp$ appears to be favored.

Radiation transport calculation methods in BNCT

International Nuclear Information System (INIS)

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

2000-01-01

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

An assessment of the feasibility of using Monte Carlo calculations to model a combined neutron/gamma electronic personal dosemeter

International Nuclear Information System (INIS)

Tanner, J.E.; Witts, D.; Tanner, R.J.; Bartlett, D.T.; Burgess, P.H.; Edwards, A.A.; More, B.R.

1995-01-01

A Monte Carlo facility has been developed for modelling the response of semiconductor devices to mixed neutron-photon fields. This utilises the code MCNP for neutron and photon transport and a new code, STRUGGLE, which has been developed to model the secondary charged particle transport. It is thus possible to predict the pulse height distribution expected from prototype electronic personal detectors, given the detector efficiency factor. Initial calculations have been performed on a simple passivated implanted planar silicon detector. This device has also been irradiated in neutron, gamma and X ray fields to verify the accuracy of the predictions. Good agreement was found between experiment and calculation. (author)

EGSNRC Monte Carlo study of the effect of photon energy and field margin in phantoms simulating small lung lesions

International Nuclear Information System (INIS)

Osei, E.K.; Darko, J.; Mosseri, A.; Jezioranski, J.

2003-01-01

The dose distribution in small lung tumors (coin lesions) is affected by the combined effects of reduced attenuation of photons and extended range of electrons in lung. The increased range of electrons in low-density tissues can lead to loss of field flatness and increased penumbra width, especially at high energies. The EGSNRC Monte Carlo code, together with DOSXYZNRC, a three-dimensional voxel dose calculation module has been used to study the characteristics of the penumbra in the region of the target-lung interfaces for various radiation beam energies, lung densities, target-field edge distances, target size, and depth. The Monte Carlo model was validated by film measurements made in acrylic (simulating a tumor) imbedded in cork (simulating the lung). Beam profiles that are deemed to be acceptable are defined as those in which no point within the planning target volume (target volume plus 1 cm margin) received less than 95% of the dose prescribed to the center of the target. For parallel opposed beams and 2 cm cube target size, 6 MV photons produce superior dose distribution with respect to penumbra at the lateral, anterior, and posterior surfaces and midplane of the simulated target, with a target-field edge distance of 2.5 cm. A lesser target-field edge distance of 2.0 cm is required for 4 MV photons to produce acceptable dose distribution. To achieve equivalent dose distribution with 10 and 18 MV photons, a target-field edge distance of 3.0 and 3.5 cm, respectively, is required. For a simulated target size of 4 cm cube, a target-field edge distance of 2, 2.5, and 3 cm is required for 6, 10, and 18 MV photons, respectively, to yield acceptable PTV coverage. The effect, which is predominant in determining the target dose, depends on the beam energy, target-field edge distance, lung density, and the depth and size of the target

A Monte Carlo Ray Tracing Model to Improve Simulations of Solar-Induced Chlorophyll Fluorescence Radiative Transfer

Science.gov (United States)

Halubok, M.; Gu, L.; Yang, Z. L.

2017-12-01

A model of light transport in a three-dimensional vegetation canopy is being designed and evaluated. The model employs Monte Carlo ray tracing technique which offers simple yet rigorous approach of quantifying the photon transport in a plant canopy. This method involves simulation of a chain of scattering and absorption events incurred by a photon on its path from the light source. Implementation of weighting mechanism helps avoid `all-or-nothing' type of interaction between a photon packet and a canopy element, i.e. at each interaction a photon packet is split into three parts, namely, reflected, transmitted and absorbed, instead of assuming complete absorption, reflection or transmission. Canopy scenes in the model are represented by a number of polygons with specified set of reflectances and transmittances. The performance of the model is being evaluated through comparison against established plant canopy reflectance models, such as 3D Radiosity-Graphics combined model which calculates bidirectional reflectance distribution function of a 3D canopy scene. This photon transport model is to be coupled to a leaf level solar-induced chlorophyll fluorescence (SIF) model with the aim of further advancing of accuracy of the modeled SIF, which, in its turn, has a potential of improving our predictive capability of terrestrial carbon uptake.

Verification of Monte Carlo transport codes by activation experiments

OpenAIRE

Chetvertkova, Vera

2013-01-01

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

Progress on RMC: a Monte Carlo neutron transport code for reactor analysis

International Nuclear Information System (INIS)

Wang, Kan; Li, Zeguang; She, Ding; Liu, Yuxuan; Xu, Qi; Shen, Huayun; Yu, Ganglin

2011-01-01

This paper presents a new 3-D Monte Carlo neutron transport code named RMC (Reactor Monte Carlo code), specifically intended for reactor physics analysis. This code is being developed by Department of Engineering Physics in Tsinghua University and written in C++ and Fortran 90 language with the latest version of RMC 2.5.0. The RMC code uses the method known as the delta-tracking method to simulate neutron transport, the advantages of which include fast simulation in complex geometries and relatively simple handling of complicated geometrical objects. Some other techniques such as computational-expense oriented method and hash-table method have been developed and implemented in RMC to speedup the calculation. To meet the requirements of reactor analysis, the RMC code has the calculational functions including criticality calculation, burnup calculation and also kinetics simulation. In this paper, comparison calculations of criticality problems, burnup problems and transient problems are carried out using RMC code and other Monte Carlo codes, and the results show that RMC performs quite well in these kinds of problems. Based on MPI, RMC succeeds in parallel computation and represents a high speed-up. This code is still under intensive development and the further work directions are mentioned at the end of this paper. (author)

SU-C-BRC-03: Development of a Novel Strategy for On-Demand Monte Carlo and Deterministic Dose Calculation Treatment Planning and Optimization for External Beam Photon and Particle Therapy

Energy Technology Data Exchange (ETDEWEB)

Yang, Y M; Bush, K; Han, B; Xing, L [Department of Radiation Oncology, Stanford University School of Medicine, Stanford, CA (United States)

2016-06-15

Purpose: Accurate and fast dose calculation is a prerequisite of precision radiation therapy in modern photon and particle therapy. While Monte Carlo (MC) dose calculation provides high dosimetric accuracy, the drastically increased computational time hinders its routine use. Deterministic dose calculation methods are fast, but problematic in the presence of tissue density inhomogeneity. We leverage the useful features of deterministic methods and MC to develop a hybrid dose calculation platform with autonomous utilization of MC and deterministic calculation depending on the local geometry, for optimal accuracy and speed. Methods: Our platform utilizes a Geant4 based “localized Monte Carlo” (LMC) method that isolates MC dose calculations only to volumes that have potential for dosimetric inaccuracy. In our approach, additional structures are created encompassing heterogeneous volumes. Deterministic methods calculate dose and energy fluence up to the volume surfaces, where the energy fluence distribution is sampled into discrete histories and transported using MC. Histories exiting the volume are converted back into energy fluence, and transported deterministically. By matching boundary conditions at both interfaces, deterministic dose calculation account for dose perturbations “downstream” of localized heterogeneities. Hybrid dose calculation was performed for water and anthropomorphic phantoms. Results: We achieved <1% agreement between deterministic and MC calculations in the water benchmark for photon and proton beams, and dose differences of 2%–15% could be observed in heterogeneous phantoms. The saving in computational time (a factor ∼4–7 compared to a full Monte Carlo dose calculation) was found to be approximately proportional to the volume of the heterogeneous region. Conclusion: Our hybrid dose calculation approach takes advantage of the computational efficiency of deterministic method and accuracy of MC, providing a practical tool for high

SU-C-BRC-03: Development of a Novel Strategy for On-Demand Monte Carlo and Deterministic Dose Calculation Treatment Planning and Optimization for External Beam Photon and Particle Therapy

International Nuclear Information System (INIS)

Yang, Y M; Bush, K; Han, B; Xing, L

2016-01-01

Purpose: Accurate and fast dose calculation is a prerequisite of precision radiation therapy in modern photon and particle therapy. While Monte Carlo (MC) dose calculation provides high dosimetric accuracy, the drastically increased computational time hinders its routine use. Deterministic dose calculation methods are fast, but problematic in the presence of tissue density inhomogeneity. We leverage the useful features of deterministic methods and MC to develop a hybrid dose calculation platform with autonomous utilization of MC and deterministic calculation depending on the local geometry, for optimal accuracy and speed. Methods: Our platform utilizes a Geant4 based “localized Monte Carlo” (LMC) method that isolates MC dose calculations only to volumes that have potential for dosimetric inaccuracy. In our approach, additional structures are created encompassing heterogeneous volumes. Deterministic methods calculate dose and energy fluence up to the volume surfaces, where the energy fluence distribution is sampled into discrete histories and transported using MC. Histories exiting the volume are converted back into energy fluence, and transported deterministically. By matching boundary conditions at both interfaces, deterministic dose calculation account for dose perturbations “downstream” of localized heterogeneities. Hybrid dose calculation was performed for water and anthropomorphic phantoms. Results: We achieved <1% agreement between deterministic and MC calculations in the water benchmark for photon and proton beams, and dose differences of 2%–15% could be observed in heterogeneous phantoms. The saving in computational time (a factor ∼4–7 compared to a full Monte Carlo dose calculation) was found to be approximately proportional to the volume of the heterogeneous region. Conclusion: Our hybrid dose calculation approach takes advantage of the computational efficiency of deterministic method and accuracy of MC, providing a practical tool for high

Relativistic theory of particles in a scattering flow III: photon transport.

Science.gov (United States)

Achterberg, A.; Norman, C. A.

2018-06-01

We use the theory developed in Achterberg & Norman (2018a) and Achterberg & Norman (2018b) to calculate the stress due to photons that are scattered elastically by a relativistic flow. We show that the energy-momentum tensor of the radiation takes the form proposed by Eckart (1940). In particular we show that no terms associated with a bulk viscosity appear if one makes the diffusion approximation for radiation transport and treats the radiation as a separate fluid. We find only shear (dynamic) viscosity terms and heat flow terms in our expression for the energy-momentum tensor. This conclusion holds quite generally for different forms of scattering: Krook-type integral scattering, diffusive (Fokker-Planck) scattering and Thomson scattering. We also derive the transport equation in the diffusion approximation that shows the effects of the flow on the photon gas in the form of a combination of adiabatic heating and an irreversible heating term. We find no diffusive changes to the comoving number density and energy density of the scattered photons, in contrast with some published results in Radiation Hydrodynamics. It is demonstrated that these diffusive corrections to the number- and energy density of the photons are in fact higher-order terms that can (and should) be neglected in the diffusion approximation. Our approach eliminates these terms at the root of the expansion that yields the anisotropic terms in the phase-space density of particles and photons, the terms responsible for the photon viscosity.

Data decomposition of Monte Carlo particle transport simulations via tally servers

International Nuclear Information System (INIS)

Romano, Paul K.; Siegel, Andrew R.; Forget, Benoit; Smith, Kord

2013-01-01

An algorithm for decomposing large tally data in Monte Carlo particle transport simulations is developed, analyzed, and implemented in a continuous-energy Monte Carlo code, OpenMC. The algorithm is based on a non-overlapping decomposition of compute nodes into tracking processors and tally servers. The former are used to simulate the movement of particles through the domain while the latter continuously receive and update tally data. A performance model for this approach is developed, suggesting that, for a range of parameters relevant to LWR analysis, the tally server algorithm should perform with minimal overhead on contemporary supercomputers. An implementation of the algorithm in OpenMC is then tested on the Intrepid and Titan supercomputers, supporting the key predictions of the model over a wide range of parameters. We thus conclude that the tally server algorithm is a successful approach to circumventing classical on-node memory constraints en route to unprecedentedly detailed Monte Carlo reactor simulations

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.

A Study on Efficiency Improvement of the Hybrid Monte Carlo/Deterministic Method for Global Transport Problems

International Nuclear Information System (INIS)

Kim, Jong Woo; Woo, Myeong Hyeon; Kim, Jae Hyun; Kim, Do Hyun; Shin, Chang Ho; Kim, Jong Kyung

2017-01-01

In this study hybrid Monte Carlo/Deterministic method is explained for radiation transport analysis in global system. FW-CADIS methodology construct the weight window parameter and it useful at most global MC calculation. However, Due to the assumption that a particle is scored at a tally, less particles are transported to the periphery of mesh tallies. For compensation this space-dependency, we modified the module in the ADVANTG code to add the proposed method. We solved the simple test problem for comparing with result from FW-CADIS methodology, it was confirmed that a uniform statistical error was secured as intended. In the future, it will be added more practical problems. It might be useful to perform radiation transport analysis using the Hybrid Monte Carlo/Deterministic method in global transport problems.

Monte Carlo particle simulation and finite-element techniques for tandem mirror transport

International Nuclear Information System (INIS)

Rognlien, T.D.; Cohen, B.I.; Matsuda, Y.; Stewart, J.J. Jr.

1987-01-01

A description is given of numerical methods used in the study of axial transport in tandem mirrors owing to Coulomb collisions and rf diffusion. The methods are Monte Carlo particle simulations and direct solution to the Fokker-Planck equations by finite-element expansion. (author)

Monte Carlo particle simulation and finite-element techniques for tandem mirror transport

International Nuclear Information System (INIS)

Rognlien, T.D.; Cohen, B.I.; Matsuda, Y.; Stewart, J.J. Jr.

1985-12-01

A description is given of numerical methods used in the study of axial transport in tandem mirrors owing to Coulomb collisions and rf diffusion. The methods are Monte Carlo particle simulations and direct solution to the Fokker-Planck equations by finite-element expansion. 11 refs

Qualification of a Monte Carlo model of photon beams of a Lilac Elekta Precise

International Nuclear Information System (INIS)

Linares R, H. M.; Laguardia, R. A.; Lara M, E.

2014-08-01

For the simulation of the accelerator head the parameters determination that characterize the electrons primary beam that affect in the target is a step that involves a fundamental role in the precision of the Monte Carlo calculations. Applying the proposed methodology by Pena et al. [2007], in this work was carried out the qualification of the photon beams (6 MV and 15 MV) of an accelerator Elekta Precise, using the Monte Carlo code EGSnrc. The influence exerted by the characteristics of the electrons primary beam on the distribution of absorbed dose for the two energy of this equipment was studied. Using different mid energy combinations and FWHM of the electrons primary beam was calculated the dose deposited in a segmented water mannequin with its surface to 100 cm of the source. Starting from the deposited dose in the mannequin the dose curves in depth and dose profiles to different depths were built. These curves were compared with measured values in a similar experimental arrangement to the carried out simulation, applying acceptability criteria based on confidence intervals [Venselaar et al. 2001]. The dose profiles for small fields were like it was expected, to be strongly influenced by the radial distribution (FWHM). The energy/FWHM combinations that better reproduce the experimental curves of each photon beam were determined. One time determined the best combination (5.75 MeV/2 mm and 11.25 MeV/2 mm, respectively) was used for the generation of the phase spaces and the field factors calculation. A good correspondence was obtained between the simulations and the measurements for a wide range of field sizes, as well as for different types of detectors, being all the results inside of the tolerance margins. (author)

Monte-Carlo estimation of the inflight performance of the GEMS satellite x-ray polarimeter

Science.gov (United States)

Kitaguchi, Takao; Tamagawa, Toru; Hayato, Asami; Enoto, Teruaki; Yoshikawa, Akifumi; Kaneko, Kenta; Takeuchi, Yoko; Black, Kevin; Hill, Joanne; Jahoda, Keith; Krizmanic, John; Sturner, Steven; Griffiths, Scott; Kaaret, Philip; Marlowe, Hannah

2014-07-01

We report a Monte-Carlo estimation of the in-orbit performance of a cosmic X-ray polarimeter designed to be installed on the focal plane of a small satellite. The simulation uses GEANT for the transport of photons and energetic particles and results from Magboltz for the transport of secondary electrons in the detector gas. We validated the simulation by comparing spectra and modulation curves with actual data taken with radioactive sources and an X-ray generator. We also estimated the in-orbit background induced by cosmic radiation in low Earth orbit.

MC++: A parallel, portable, Monte Carlo neutron transport code in C++

International Nuclear Information System (INIS)

Lee, S.R.; Cummings, J.C.; Nolen, S.D.

1997-01-01

MC++ is an implicit multi-group Monte Carlo neutron transport code written in C++ and based on the Parallel Object-Oriented Methods and Applications (POOMA) class library. MC++ runs in parallel on and is portable to a wide variety of platforms, including MPPs, SMPs, and clusters of UNIX workstations. MC++ is being developed to provide transport capabilities to the Accelerated Strategic Computing Initiative (ASCI). It is also intended to form the basis of the first transport physics framework (TPF), which is a C++ class library containing appropriate abstractions, objects, and methods for the particle transport problem. The transport problem is briefly described, as well as the current status and algorithms in MC++ for solving the transport equation. The alpha version of the POOMA class library is also discussed, along with the implementation of the transport solution algorithms using POOMA. Finally, a simple test problem is defined and performance and physics results from this problem are discussed on a variety of platforms

Liquid argon as an electron/photon detector in the energy range of 50 MeV to 2 GeV: a Monte Carlo investigation

International Nuclear Information System (INIS)

Goodman, M.S.; Denis, G.; Hall, M.; Karpovsky, A.; Wilson, R.; Gabriel, T.A.; Bishop, B.L.

1980-12-01

Monte Carlo techniques which have been used to study the characteristics of a proposed electron/photon detector based on the total absorption of electromagnetic showers in liquid argon have been investigated. The energy range studied was 50 MeV to 2 GeV. Results are presented on the energy and angular resolution predicted for the device, along with the detailed predictions of the transverse and longitudinal shower distributions. Comparisons are made with other photon detectors, and possible applications are discussed

Benchmark of PENELOPE code for low-energy photon transport: dose comparisons with MCNP4 and EGS4

International Nuclear Information System (INIS)

Ye, Sung-Joon; Brezovich, Ivan A; Pareek, Prem; Naqvi, Shahid A

2004-01-01

The expanding clinical use of low-energy photon emitting 125 I and 103 Pd seeds in recent years has led to renewed interest in their dosimetric properties. Numerous papers pointed out that higher accuracy could be obtained in Monte Carlo simulations by utilizing newer libraries for the low-energy photon cross-sections, such as XCOM and EPDL97. The recently developed PENELOPE 2001 Monte Carlo code is user friendly and incorporates photon cross-section data from the EPDL97. The code has been verified for clinical dosimetry of high-energy electron and photon beams, but has not yet been tested at low energies. In the present work, we have benchmarked the PENELOPE code for 10-150 keV photons. We computed radial dose distributions from 0 to 10 cm in water at photon energies of 10-150 keV using both PENELOPE and MCNP4C with either DLC-146 or DLC-200 cross-section libraries, assuming a point source located at the centre of a 30 cm diameter and 20 cm length cylinder. Throughout the energy range of simulated photons (except for 10 keV), PENELOPE agreed within statistical uncertainties (at worst ±5%) with MCNP/DLC-146 in the entire region of 1-10 cm and with published EGS4 data up to 5 cm. The dose at 1 cm (or dose rate constant) of PENELOPE agreed with MCNP/DLC-146 and EGS4 data within approximately ±2% in the range of 20-150 keV, while MCNP/DLC-200 produced values up to 9% lower in the range of 20-100 keV than PENELOPE or the other codes. However, the differences among the four datasets became negligible above 100 keV

Benchmark of PENELOPE code for low-energy photon transport: dose comparisons with MCNP4 and EGS4.

Science.gov (United States)

Ye, Sung-Joon; Brezovich, Ivan A; Pareek, Prem; Naqvi, Shahid A

2004-02-07

The expanding clinical use of low-energy photon emitting 125I and 103Pd seeds in recent years has led to renewed interest in their dosimetric properties. Numerous papers pointed out that higher accuracy could be obtained in Monte Carlo simulations by utilizing newer libraries for the low-energy photon cross-sections, such as XCOM and EPDL97. The recently developed PENELOPE 2001 Monte Carlo code is user friendly and incorporates photon cross-section data from the EPDL97. The code has been verified for clinical dosimetry of high-energy electron and photon beams, but has not yet been tested at low energies. In the present work, we have benchmarked the PENELOPE code for 10-150 keV photons. We computed radial dose distributions from 0 to 10 cm in water at photon energies of 10-150 keV using both PENELOPE and MCNP4C with either DLC-146 or DLC-200 cross-section libraries, assuming a point source located at the centre of a 30 cm diameter and 20 cm length cylinder. Throughout the energy range of simulated photons (except for 10 keV), PENELOPE agreed within statistical uncertainties (at worst +/- 5%) with MCNP/DLC-146 in the entire region of 1-10 cm and with published EGS4 data up to 5 cm. The dose at 1 cm (or dose rate constant) of PENELOPE agreed with MCNP/DLC-146 and EGS4 data within approximately +/- 2% in the range of 20-150 keV, while MCNP/DLC-200 produced values up to 9% lower in the range of 20-100 keV than PENELOPE or the other codes. However, the differences among the four datasets became negligible above 100 keV.

A Path Space Extension for Robust Light Transport Simulation

DEFF Research Database (Denmark)

Hachisuka, Toshiya; Pantaleoni, Jacopo; Jensen, Henrik Wann

2012-01-01

We present a new sampling space for light transport paths that makes it possible to describe Monte Carlo path integration and photon density estimation in the same framework. A key contribution of our paper is the introduction of vertex perturbations, which extends the space of paths with loosely...

Energy deposition model for I-125 photon radiation in water

International Nuclear Information System (INIS)

Fuss, M.C.; Garcia, G.; Munoz, A.; Oller, J.C.; Blanco, F.; Limao-Vieira, P.; Williart, A.; Garcia, G.; Huerga, C.; Tellez, M.

2010-01-01

In this study, an electron-tracking Monte Carlo algorithm developed by us is combined with established photon transport models in order to simulate all primary and secondary particle interactions in water for incident photon radiation. As input parameters for secondary electron interactions, electron scattering cross sections by water molecules and experimental energy loss spectra are used. With this simulation, the resulting energy deposition can be modelled at the molecular level, yielding detailed information about localization and type of single collision events. The experimental emission spectrum of I-125 seeds, as used for radiotherapy of different tumours, was used for studying the energy deposition in water when irradiating with this radionuclide. (authors)

Energy deposition model for I-125 photon radiation in water

Energy Technology Data Exchange (ETDEWEB)

Fuss, M.C.; Garcia, G. [Instituto de Fisica Fundamental, Consejo Superior de Investigaciones Cientificas (CSIC), Madrid (Spain); Munoz, A.; Oller, J.C. [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT), Madrid (Spain); Blanco, F. [Departamento de Fisica Atomica, Molecular y Nuclear, Universidad Complutense de Madrid (Spain); Limao-Vieira, P. [Laboratorio de Colisoes Atomicas e Moleculares, Departamento de Fisica, CEFITEC, FCT-Universidade Nova de Lisboa, Caparica (Portugal); Williart, A.; Garcia, G. [Departamento de Fisica de los Materiales, Universidad Nacional de Educacion a Distancia, Madrid (Spain); Huerga, C.; Tellez, M. [Hospital Universitario La Paz, Madrid (Spain)

2010-10-15

In this study, an electron-tracking Monte Carlo algorithm developed by us is combined with established photon transport models in order to simulate all primary and secondary particle interactions in water for incident photon radiation. As input parameters for secondary electron interactions, electron scattering cross sections by water molecules and experimental energy loss spectra are used. With this simulation, the resulting energy deposition can be modelled at the molecular level, yielding detailed information about localization and type of single collision events. The experimental emission spectrum of I-125 seeds, as used for radiotherapy of different tumours, was used for studying the energy deposition in water when irradiating with this radionuclide. (authors)

MCNP trademark Monte Carlo: A precis of MCNP

International Nuclear Information System (INIS)

Adams, K.J.

1996-01-01

MCNP trademark is a general purpose three-dimensional time-dependent neutron, photon, and electron transport code. It is highly portable and user-oriented, and backed by stringent software quality assurance practices and extensive experimental benchmarks. The cross section database is based upon the best evaluations available. MCNP incorporates state-of-the-art analog and adaptive Monte Carlo techniques. The code is documented in a 600 page manual which is augmented by numerous Los Alamos technical reports which detail various aspects of the code. MCNP represents over a megahour of development and refinement over the past 50 years and an ongoing commitment to excellence

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

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

Dose calculation methods in photon beam therapy using energy deposition kernels

International Nuclear Information System (INIS)

Ahnesjoe, A.

1991-01-01

The problem of calculating accurate dose distributions in treatment planning of megavoltage photon radiation therapy has been studied. New dose calculation algorithms using energy deposition kernels have been developed. The kernels describe the transfer of energy by secondary particles from a primary photon interaction site to its surroundings. Monte Carlo simulations of particle transport have been used for derivation of kernels for primary photon energies form 0.1 MeV to 50 MeV. The trade off between accuracy and calculational speed has been addressed by the development of two algorithms; one point oriented with low computional overhead for interactive use and one for fast and accurate calculation of dose distributions in a 3-dimensional lattice. The latter algorithm models secondary particle transport in heterogeneous tissue by scaling energy deposition kernels with the electron density of the tissue. The accuracy of the methods has been tested using full Monte Carlo simulations for different geometries, and found to be superior to conventional algorithms based on scaling of broad beam dose distributions. Methods have also been developed for characterization of clinical photon beams in entities appropriate for kernel based calculation models. By approximating the spectrum as laterally invariant, an effective spectrum and dose distribution for contaminating charge particles are derived form depth dose distributions measured in water, using analytical constraints. The spectrum is used to calculate kernels by superposition of monoenergetic kernels. The lateral energy fluence distribution is determined by deconvolving measured lateral dose distributions by a corresponding pencil beam kernel. Dose distributions for contaminating photons are described using two different methods, one for estimation of the dose outside of the collimated beam, and the other for calibration of output factors derived from kernel based dose calculations. (au)

MCMEG: Simulations of both PDD and TPR for 6 MV LINAC photon beam using different MC codes

International Nuclear Information System (INIS)

Fonseca, T.C.F.; Mendes, B.M.; Lacerda, M.A.S.; Silva, L.A.C.; Paixão, L.

2017-01-01

The Monte Carlo Modelling Expert Group (MCMEG) is an expert network specializing in Monte Carlo radiation transport and the modelling and simulation applied to the radiation protection and dosimetry research field. For the first inter-comparison task the group launched an exercise to model and simulate a 6 MV LINAC photon beam using the Monte Carlo codes available within their laboratories and validate their simulated results by comparing them with experimental measurements carried out in the National Cancer Institute (INCA) in Rio de Janeiro, Brazil. The experimental measurements were performed using an ionization chamber with calibration traceable to a Secondary Standard Dosimetry Laboratory (SSDL). The detector was immersed in a water phantom at different depths and was irradiated with a radiation field size of 10×10 cm 2 . This exposure setup was used to determine the dosimetric parameters Percentage Depth Dose (PDD) and Tissue Phantom Ratio (TPR). The validation process compares the MC calculated results to the experimental measured PDD20,10 and TPR20,10. Simulations were performed reproducing the experimental TPR20,10 quality index which provides a satisfactory description of both the PDD curve and the transverse profiles at the two depths measured. This paper reports in detail the modelling process using MCNPx, MCNP6, EGSnrc and Penelope Monte Carlo codes, the source and tally descriptions, the validation processes and the results. - Highlights: • MCMEG is an expert network specializing in Monte Carlo radiation transport. • MCNPx, MCNP6, EGSnrc and Penelope Monte Carlo codes are used. • Exercise to model and simulate a 6 MV LINAC photon beam using the Monte Carlo codes. • The PDD 20,10 and TPR 20,10 dosimetric parameters were compared with real data. • The paper reports in the modelling process using different Monte Carlo codes.

The Monte Carlo applied for calculation dose

International Nuclear Information System (INIS)

Peixoto, J.E.

1988-01-01

The Monte Carlo method is showed for the calculation of absorbed dose. The trajectory of the photon is traced simulating sucessive interaction between the photon and the substance that consist the human body simulator. The energy deposition in each interaction of the simulator organ or tissue per photon is also calculated. (C.G.C.) [pt

Development of a photon-cell interactive monte carlo simulation for non-invasive measurement of blood glucose level by Raman spectroscopy.

Science.gov (United States)

Sakota, Daisuke; Kosaka, Ryo; Nishida, Masahiro; Maruyama, Osamu

2015-01-01

Turbidity variation is one of the major limitations in Raman spectroscopy for quantifying blood components, such as glucose, non-invasively. To overcome this limitation, we have developed a Raman scattering simulation using a photon-cell interactive Monte Carlo (pciMC) model that tracks photon migration in both the extra- and intracellular spaces without relying on the macroscopic scattering phase function and anisotropy factor. The interaction of photons at the plasma-cell boundary of randomly oriented three-dimensionally biconcave red blood cells (RBCs) is modeled using geometric optics. The validity of the developed pciMCRaman was investigated by comparing simulation and experimental results of Raman spectroscopy of glucose level in a bovine blood sample. The scattering of the excitation laser at a wavelength of 785 nm was simulated considering the changes in the refractive index of the extracellular solution. Based on the excitation laser photon distribution within the blood, the Raman photon derived from the hemoglobin and glucose molecule at the Raman shift of 1140 cm(-1) = 862 nm was generated, and the photons reaching the detection area were counted. The simulation and experimental results showed good correlation. It is speculated that pciMCRaman can provide information about the ability and limitations of the measurement of blood glucose level.

Analyzing the photonic band gaps in two-dimensional plasma photonic crystals with fractal Sierpinski gasket structure based on the Monte Carlo method

Energy Technology Data Exchange (ETDEWEB)

Zhang, Hai-Feng, E-mail: hanlor@163.com [College of Optoelectronic Engineering, Nanjing University of Posts and Telecommunications, Nanjing, 210023 ,China (China); Key Laboratory of Radar Imaging and Microwave Photonics (Nanjing Univ. Aeronaut. Astronaut.), Ministry of Education, Nanjing University of Aeronautics and Astronautics, Nanjing, 210016 (China); Liu, Shao-Bin [Key Laboratory of Radar Imaging and Microwave Photonics (Nanjing Univ. Aeronaut. Astronaut.), Ministry of Education, Nanjing University of Aeronautics and Astronautics, Nanjing, 210016 (China)

2016-08-15

In this paper, the properties of photonic band gaps (PBGs) in two types of two-dimensional plasma-dielectric photonic crystals (2D PPCs) under a transverse-magnetic (TM) wave are theoretically investigated by a modified plane wave expansion (PWE) method where Monte Carlo method is introduced. The proposed PWE method can be used to calculate the band structures of 2D PPCs which possess arbitrary-shaped filler and any lattice. The efficiency and convergence of the present method are discussed by a numerical example. The configuration of 2D PPCs is the square lattices with fractal Sierpinski gasket structure whose constituents are homogeneous and isotropic. The type-1 PPCs is filled with the dielectric cylinders in the plasma background, while its complementary structure is called type-2 PPCs, in which plasma cylinders behave as the fillers in the dielectric background. The calculated results reveal that the enough accuracy and good convergence can be obtained, if the number of random sampling points of Monte Carlo method is large enough. The band structures of two types of PPCs with different fractal orders of Sierpinski gasket structure also are theoretically computed for a comparison. It is demonstrate that the PBGs in higher frequency region are more easily produced in the type-1 PPCs rather than in the type-2 PPCs. Sierpinski gasket structure introduced in the 2D PPCs leads to a larger cutoff frequency, enhances and induces more PBGs in high frequency region. The effects of configurational parameters of two types of PPCs on the PBGs are also investigated in detail. The results show that the PBGs of the PPCs can be easily manipulated by tuning those parameters. The present type-1 PPCs are more suitable to design the tunable compacted devices.

A fully coupled Monte Carlo/discrete ordinates solution to the neutron transport equation. Final report

Energy Technology Data Exchange (ETDEWEB)

Baker, Randal Scott [Univ. of Arizona, Tucson, AZ (United States)

1990-01-01

The neutron transport equation is solved by a hybrid method that iteratively couples regions where deterministic (S_N) and stochastic (Monte Carlo) methods are applied. Unlike previous hybrid methods, the Monte Carlo and S_N regions are fully coupled in the sense that no assumption is made about geometrical separation or decoupling. The hybrid method provides a new means of solving problems involving both optically thick and optically thin regions that neither Monte Carlo nor S_N is well suited for by themselves. The fully coupled Monte Carlo/S_N technique consists of defining spatial and/or energy regions of a problem in which either a Monte Carlo calculation or an S_N calculation is to be performed. The Monte Carlo region may comprise the entire spatial region for selected energy groups, or may consist of a rectangular area that is either completely or partially embedded in an arbitrary S_N region. The Monte Carlo and S_N regions are then connected through the common angular boundary fluxes, which are determined iteratively using the response matrix technique, and volumetric sources. The hybrid method has been implemented in the S_N code TWODANT by adding special-purpose Monte Carlo subroutines to calculate the response matrices and volumetric sources, and linkage subrountines to carry out the interface flux iterations. The common angular boundary fluxes are included in the S_N code as interior boundary sources, leaving the logic for the solution of the transport flux unchanged, while, with minor modifications, the diffusion synthetic accelerator remains effective in accelerating S_N calculations. The special-purpose Monte Carlo routines used are essentially analog, with few variance reduction techniques employed. However, the routines have been successfully vectorized, with approximately a factor of five increase in speed over the non-vectorized version.

Universal dependence of the total number albedo of photons on the mean number of photon scatterings

Directory of Open Access Journals (Sweden)

Ljubenov Vladan L.

2011-01-01

Full Text Available This paper presents the results of research on photon reflection from plane targets based on Monte Carlo simulations performed by the MCNP code. Five materials (water, concrete, aluminum, iron, and copper are examined in the area of initial photon energies of up to 200 keV. The values of the total number albedo for photons dependent on the initial photon energy or the mean number of photon scatterings are calculated and graphically presented. We have shown that the values of the total number albedo for different target materials, expressed as a function of the mean number of photon scatterings, are in good agreement with each other and can be approximated by simple, universal analytic functions obtained by the least squares method. The accuracy of these analytic appoximations is confirmed by their comparison with the results of PENELOPE and FOTELP Monte Carlo codes.

Multidimensional electron-photon transport with standard discrete ordinates codes

International Nuclear Information System (INIS)

Drumm, C.R.

1997-01-01

A method is described for generating electron cross sections that are comparable with standard discrete ordinates codes without modification. There are many advantages of using an established discrete ordinates solver, e.g. immediately available adjoint capability. Coupled electron-photon transport capability is needed for many applications, including the modeling of the response of electronics components to space and man-made radiation environments. The cross sections have been successfully used in the DORT, TWODANT and TORT discrete ordinates codes. The cross sections are shown to provide accurate and efficient solutions to certain multidimensional electron-photon transport problems. The key to the method is a simultaneous solution of the continuous-slowing-down (CSD) portion and elastic-scattering portion of the scattering source by the Goudsmit-Saunderson theory. The resulting multigroup-Legendre cross sections are much smaller than the true scattering cross sections that they represent. Under certain conditions, the cross sections are guaranteed positive and converge with a low-order Legendre expansion

Multidimensional electron-photon transport with standard discrete ordinates codes

International Nuclear Information System (INIS)

Drumm, C.R.

1997-01-01

A method is described for generating electron cross sections that are compatible with standard discrete ordinates codes without modification. There are many advantages to using an established discrete ordinates solver, e.g., immediately available adjoint capability. Coupled electron-photon transport capability is needed for many applications, including the modeling of the response of electronics components to space and synthetic radiation environments. The cross sections have been successfully used in the DORT, TWODANT, and TORT discrete ordinates codes. The cross sections are shown to provide accurate and efficient solutions to certain multidimensional electron-photon transport problems. The key to the method is a simultaneous solution of the continuous-slowing-down and elastic-scattering portions of the scattering source by the Goudsmit-Saunderson theory. The resulting multigroup-Legendre cross sections are much smaller than the true scattering cross sections that they represent. Under certain conditions, the cross sections are guaranteed positive and converge with a low-order Legendre expansion

Monte Carlo methods in electron transport problems. Pt. 1

International Nuclear Information System (INIS)

Cleri, F.

1989-01-01

The condensed-history Monte Carlo method for charged particles transport is reviewed and discussed starting from a general form of the Boltzmann equation (Part I). The physics of the electronic interactions, together with some pedagogic example will be introduced in the part II. The lecture is directed to potential users of the method, for which it can be a useful introduction to the subject matter, and wants to establish the basis of the work on the computer code RECORD, which is at present in a developing stage

Application of the Monte Carlo method to diagnostic radiology

International Nuclear Information System (INIS)

Persliden, J.

1986-01-01

A Monte Carlo program for photon transport is developed. The program is used to investigate the energy imparted to water slabs (simulating patients), and the related backscattered and transmitted energies as functions of primary photon energy and water slab thickness. The accuracy of the results depends on the cross-section data for the probabilities of the various interactions in the slab and on the physical quantity calculated. Backscattered energy fractions can vary by as much as 10-20 %, using different sets of published data for the photoelectric cross section while imparted fractions are only slightly affected. The results are used to calculate improved conversion factors for determining the energy imparted to the patient in X-ray diagnostic examinations from measurements of the air collision kerma integrated over beam area. The small angle distribution of scattered photons transmitted through a water slab, relevant to problems of image quality, is calculated taking into account the diffraction phenomena of liquid water. The calculations are performed with a collision density estimator. This estimator makes it possible to calculate important physical quantities which are virtually impracticable to assess with the Monte Carlo codes commonly used in medical physics or in experiments. With the collision density estimator, the influence of air gaps on the reduction of scattered radiation is investigated for different detectors, field areas and primary X-ray spectra. Contrast degradation and contrast improvement factors are given as functions of field area for various air gaps. (With 105 refs.) (author)

Monte Carlo techniques in diagnostic and therapeutic nuclear medicine

International Nuclear Information System (INIS)

Zaidi, H.

2002-01-01

Monte Carlo techniques have become one of the most popular tools in different areas of medical radiation physics following the development and subsequent implementation of powerful computing systems for clinical use. In particular, they have been extensively applied to simulate processes involving random behaviour and to quantify physical parameters that are difficult or even impossible to calculate analytically or to determine by experimental measurements. The use of the Monte Carlo method to simulate radiation transport turned out to be the most accurate means of predicting absorbed dose distributions and other quantities of interest in the radiation treatment of cancer patients using either external or radionuclide radiotherapy. The same trend has occurred for the estimation of the absorbed dose in diagnostic procedures using radionuclides. There is broad consensus in accepting that the earliest Monte Carlo calculations in medical radiation physics were made in the area of nuclear medicine, where the technique was used for dosimetry modelling and computations. Formalism and data based on Monte Carlo calculations, developed by the Medical Internal Radiation Dose (MIRD) committee of the Society of Nuclear Medicine, were published in a series of supplements to the Journal of Nuclear Medicine, the first one being released in 1968. Some of these pamphlets made extensive use of Monte Carlo calculations to derive specific absorbed fractions for electron and photon sources uniformly distributed in organs of mathematical phantoms. Interest in Monte Carlo-based dose calculations with β-emitters has been revived with the application of radiolabelled monoclonal antibodies to radioimmunotherapy. As a consequence of this generalized use, many questions are being raised primarily about the need and potential of Monte Carlo techniques, but also about how accurate it really is, what would it take to apply it clinically and make it available widely to the medical physics

Design of sampling tools for Monte Carlo particle transport code JMCT

International Nuclear Information System (INIS)

Shangguan Danhua; Li Gang; Zhang Baoyin; Deng Li

2012-01-01

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

Monte Carlo simulation of MOSFET detectors for high-energy photon beams using the PENELOPE code

Science.gov (United States)

Panettieri, Vanessa; Amor Duch, Maria; Jornet, Núria; Ginjaume, Mercè; Carrasco, Pablo; Badal, Andreu; Ortega, Xavier; Ribas, Montserrat

2007-01-01

The aim of this work was the Monte Carlo (MC) simulation of the response of commercially available dosimeters based on metal oxide semiconductor field effect transistors (MOSFETs) for radiotherapeutic photon beams using the PENELOPE code. The studied Thomson&Nielsen TN-502-RD MOSFETs have a very small sensitive area of 0.04 mm2 and a thickness of 0.5 µm which is placed on a flat kapton base and covered by a rounded layer of black epoxy resin. The influence of different metallic and Plastic water™ build-up caps, together with the orientation of the detector have been investigated for the specific application of MOSFET detectors for entrance in vivo dosimetry. Additionally, the energy dependence of MOSFET detectors for different high-energy photon beams (with energy >1.25 MeV) has been calculated. Calculations were carried out for simulated 6 MV and 18 MV x-ray beams generated by a Varian Clinac 1800 linear accelerator, a Co-60 photon beam from a Theratron 780 unit, and monoenergetic photon beams ranging from 2 MeV to 10 MeV. The results of the validation of the simulated photon beams show that the average difference between MC results and reference data is negligible, within 0.3%. MC simulated results of the effect of the build-up caps on the MOSFET response are in good agreement with experimental measurements, within the uncertainties. In particular, for the 18 MV photon beam the response of the detectors under a tungsten cap is 48% higher than for a 2 cm Plastic water™ cap and approximately 26% higher when a brass cap is used. This effect is demonstrated to be caused by positron production in the build-up caps of higher atomic number. This work also shows that the MOSFET detectors produce a higher signal when their rounded side is facing the beam (up to 6%) and that there is a significant variation (up to 50%) in the response of the MOSFET for photon energies in the studied energy range. All the results have shown that the PENELOPE code system can

Monte Carlo simulation of MOSFET detectors for high-energy photon beams using the PENELOPE code.

Science.gov (United States)

Panettieri, Vanessa; Duch, Maria Amor; Jornet, Núria; Ginjaume, Mercè; Carrasco, Pablo; Badal, Andreu; Ortega, Xavier; Ribas, Montserrat

2007-01-07

The aim of this work was the Monte Carlo (MC) simulation of the response of commercially available dosimeters based on metal oxide semiconductor field effect transistors (MOSFETs) for radiotherapeutic photon beams using the PENELOPE code. The studied Thomson&Nielsen TN-502-RD MOSFETs have a very small sensitive area of 0.04 mm(2) and a thickness of 0.5 microm which is placed on a flat kapton base and covered by a rounded layer of black epoxy resin. The influence of different metallic and Plastic water build-up caps, together with the orientation of the detector have been investigated for the specific application of MOSFET detectors for entrance in vivo dosimetry. Additionally, the energy dependence of MOSFET detectors for different high-energy photon beams (with energy >1.25 MeV) has been calculated. Calculations were carried out for simulated 6 MV and 18 MV x-ray beams generated by a Varian Clinac 1800 linear accelerator, a Co-60 photon beam from a Theratron 780 unit, and monoenergetic photon beams ranging from 2 MeV to 10 MeV. The results of the validation of the simulated photon beams show that the average difference between MC results and reference data is negligible, within 0.3%. MC simulated results of the effect of the build-up caps on the MOSFET response are in good agreement with experimental measurements, within the uncertainties. In particular, for the 18 MV photon beam the response of the detectors under a tungsten cap is 48% higher than for a 2 cm Plastic water cap and approximately 26% higher when a brass cap is used. This effect is demonstrated to be caused by positron production in the build-up caps of higher atomic number. This work also shows that the MOSFET detectors produce a higher signal when their rounded side is facing the beam (up to 6%) and that there is a significant variation (up to 50%) in the response of the MOSFET for photon energies in the studied energy range. All the results have shown that the PENELOPE code system can successfully

Adaptive multilevel splitting for Monte Carlo particle transport

Directory of Open Access Journals (Sweden)

Louvin Henri

2017-01-01

Full Text Available In the Monte Carlo simulation of particle transport, and especially for shielding applications, variance reduction techniques are widely used to help simulate realisations of rare events and reduce the relative errors on the estimated scores for a given computation time. Adaptive Multilevel Splitting (AMS is one of these variance reduction techniques that has recently appeared in the literature. In the present paper, we propose an alternative version of the AMS algorithm, adapted for the first time to the field of particle transport. Within this context, it can be used to build an unbiased estimator of any quantity associated with particle tracks, such as flux, reaction rates or even non-Boltzmann tallies like pulse-height tallies and other spectra. Furthermore, the efficiency of the AMS algorithm is shown not to be very sensitive to variations of its input parameters, which makes it capable of significant variance reduction without requiring extended user effort.

Investigating the impossible: Monte Carlo simulations

International Nuclear Information System (INIS)

Kramer, Gary H.; Crowley, Paul; Burns, Linda C.

2000-01-01

Designing and testing new equipment can be an expensive and time consuming process or the desired performance characteristics may preclude its construction due to technological shortcomings. Cost may also prevent equipment being purchased for other scenarios to be tested. An alternative is to use Monte Carlo simulations to make the investigations. This presentation exemplifies how Monte Carlo code calculations can be used to fill the gap. An example is given for the investigation of two sizes of germanium detector (70 mm and 80 mm diameter) at four different crystal thicknesses (15, 20, 25, and 30 mm) and makes predictions on how the size affects the counting efficiency and the Minimum Detectable Activity (MDA). The Monte Carlo simulations have shown that detector efficiencies can be adequately modelled using photon transport if the data is used to investigate trends. The investigation of the effect of detector thickness on the counting efficiency has shown that thickness for a fixed diameter detector of either 70 mm or 80 mm is unimportant up to 60 keV. At higher photon energies, the counting efficiency begins to decrease as the thickness decreases as expected. The simulations predict that the MDA of either the 70 mm or 80 mm diameter detectors does not differ by more than a factor of 1.15 at 17 keV or 1.2 at 60 keV when comparing detectors of equivalent thicknesses. The MDA is slightly increased at 17 keV, and rises by about 52% at 660 keV, when the thickness is decreased from 30 mm to 15 mm. One could conclude from this information that the extra cost associated with the larger area Ge detectors may not be justified for the slight improvement predicted in the MDA. (author)

Monte Carlo simulations used to calculate the energy deposited in the coronary artery lumen as a function of iodine concentration and photon energy.

Science.gov (United States)

Hocine, Nora; Meignan, Michel; Masset, Hélène

2018-04-01

To better understand the risks of cumulative medical X-ray investigations and the possible causal role of contrast agent on the coronary artery wall, the correlation between iodinated contrast media and the increase of energy deposited in the coronary artery lumen as a function of iodine concentration and photon energy is investigated. The calculations of energy deposition have been performed using Monte Carlo (MC) simulation codes, namely PENetration and Energy LOss of Positrons and Electrons (PENELOPE) and Monte Carlo N-Particle eXtended (MCNPX). Exposure of a cylinder phantom, artery and a metal stent (AISI 316L) to several X-ray photon beams were simulated. For the energies used in cardiac imaging the energy deposited in the coronary artery lumen increases with the quantity of iodine. Monte Carlo calculations indicate a strong dependence of the energy enhancement factor (EEF) on photon energy and iodine concentration. The maximum value of EEF is equal to 25; this factor is showed for 83 keV and for 400 mg Iodine/mL. No significant impact of the stent is observed on the absorbed dose in the artery for incident X-ray beams with mean energies of 44, 48, 52 and 55 keV. A strong correlation was shown between the increase in the concentration of iodine and the energy deposited in the coronary artery lumen for the energies used in cardiac imaging and over the energy range between 44 and 55 keV. The data provided by this study could be useful for creating new medical imaging protocols to obtain better diagnostic information with a lower level of radiation exposure.

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

The Monte Carlo event generator DPMJET-III

International Nuclear Information System (INIS)

Roesler, S.; Engel, R.

2001-01-01

A new version of the Monte Carlo event generator DPMJET is presented. It is a code system based on the Dual Parton Model and unifies all features of the DTUNUC-2, DPMJET-II and PHOJET1.12 event generators. DPMJET-III allows the simulation of hadron-hadron, hadron-nucleus, nucleus-nucleus, photon-hadron, photon-photon and photon-nucleus interactions from a few GeV up to the highest cosmic ray energies. (orig.)

Automatic modeling for the Monte Carlo transport code Geant4

International Nuclear Information System (INIS)

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

2015-01-01

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

A parallel Monte Carlo code for planar and SPECT imaging: implementation, verification and applications in (131)I SPECT.

Science.gov (United States)

Dewaraja, Yuni K; Ljungberg, Michael; Majumdar, Amitava; Bose, Abhijit; Koral, Kenneth F

2002-02-01

This paper reports the implementation of the SIMIND Monte Carlo code on an IBM SP2 distributed memory parallel computer. Basic aspects of running Monte Carlo particle transport calculations on parallel architectures are described. Our parallelization is based on equally partitioning photons among the processors and uses the Message Passing Interface (MPI) library for interprocessor communication and the Scalable Parallel Random Number Generator (SPRNG) to generate uncorrelated random number streams. These parallelization techniques are also applicable to other distributed memory architectures. A linear increase in computing speed with the number of processors is demonstrated for up to 32 processors. This speed-up is especially significant in Single Photon Emission Computed Tomography (SPECT) simulations involving higher energy photon emitters, where explicit modeling of the phantom and collimator is required. For (131)I, the accuracy of the parallel code is demonstrated by comparing simulated and experimental SPECT images from a heart/thorax phantom. Clinically realistic SPECT simulations using the voxel-man phantom are carried out to assess scatter and attenuation correction.

MONTE CARLO SIMULATION MODEL OF ENERGETIC PROTON TRANSPORT THROUGH SELF-GENERATED ALFVEN WAVES

Energy Technology Data Exchange (ETDEWEB)

Afanasiev, A.; Vainio, R., E-mail: alexandr.afanasiev@helsinki.fi [Department of Physics, University of Helsinki (Finland)

2013-08-15

A new Monte Carlo simulation model for the transport of energetic protons through self-generated Alfven waves is presented. The key point of the model is that, unlike the previous ones, it employs the full form (i.e., includes the dependence on the pitch-angle cosine) of the resonance condition governing the scattering of particles off Alfven waves-the process that approximates the wave-particle interactions in the framework of quasilinear theory. This allows us to model the wave-particle interactions in weak turbulence more adequately, in particular, to implement anisotropic particle scattering instead of isotropic scattering, which the previous Monte Carlo models were based on. The developed model is applied to study the transport of flare-accelerated protons in an open magnetic flux tube. Simulation results for the transport of monoenergetic protons through the spectrum of Alfven waves reveal that the anisotropic scattering leads to spatially more distributed wave growth than isotropic scattering. This result can have important implications for diffusive shock acceleration, e.g., affect the scattering mean free path of the accelerated particles in and the size of the foreshock region.

Analytical model of the binary multileaf collimator of tomotherapy for Monte Carlo simulations

International Nuclear Information System (INIS)

Sterpin, E; Vynckier, S; Salvat, F; Olivera, G H

2008-01-01

Helical Tomotherapy (HT) delivers intensity-modulated radiotherapy by the means of many configurations of the binary multi-leaf collimator (MLC). The aim of the present study was to devise a method, which we call the 'transfer function' (TF) method, to perform the transport of particles through the MLC much faster than the time consuming Monte Carlo (MC) simulation and with no significant loss of accuracy. The TF method consists of calculating, for each photon in the phase-space file, the attenuation factor for each leaf (up to three) that the photon passes, assuming straight propagation through closed leaves, and storing these factors in a modified phase-space file. To account for the transport through the MLC in a given configuration, the weight of a photon is simply multiplied by the attenuation factors of the leaves that are intersected by the photon ray and are closed. The TF method was combined with the PENELOPE MC code, and validated with measurements for the three static field sizes available (40x5, 40x2.5 and 40x1 cm 2 ) and for some MLC patterns. The TF method allows a large reduction in computation time, without introducing appreciable deviations from the result of full MC simulations

Multidimensional electron-photon transport with standard discrete ordinates codes

International Nuclear Information System (INIS)

Drumm, C.R.

1995-01-01

A method is described for generating electron cross sections that are compatible with standard discrete ordinates codes without modification. There are many advantages of using an established discrete ordinates solver, e.g. immediately available adjoint capability. Coupled electron-photon transport capability is needed for many applications, including the modeling of the response of electronics components to space and man-made radiation environments. The cross sections have been successfully used in the DORT, TWODANT and TORT discrete ordinates codes. The cross sections are shown to provide accurate and efficient solutions to certain multidimensional electronphoton transport problems

Applications of Monte Carlo method in Medical Physics

International Nuclear Information System (INIS)

Diez Rios, A.; Labajos, M.

1989-01-01

The basic ideas of Monte Carlo techniques are presented. Random numbers and their generation by congruential methods, which underlie Monte Carlo calculations are shown. Monte Carlo techniques to solve integrals are discussed. The evaluation of a simple monodimensional integral with a known answer, by means of two different Monte Carlo approaches are discussed. The basic principles to simualate on a computer photon histories reduce variance and the current applications in Medical Physics are commented. (Author)

Monte Carlo simulations of plutonium gamma-ray spectra

International Nuclear Information System (INIS)

Koenig, Z.M.; Carlson, J.B.; Wang, Tzu-Fang; Ruhter, W.D.

1993-01-01

Monte Carlo calculations were investigated as a means of simulating the gamma-ray spectra of Pu. These simulated spectra will be used to develop and evaluate gamma-ray analysis techniques for various nondestructive measurements. Simulated spectra of calculational standards can be used for code intercomparisons, to understand systematic biases and to estimate minimum detection levels of existing and proposed nondestructive analysis instruments. The capability to simulate gamma-ray spectra from HPGe detectors could significantly reduce the costs of preparing large numbers of real reference materials. MCNP was used for the Monte Carlo transport of the photons. Results from the MCNP calculations were folded in with a detector response function for a realistic spectrum. Plutonium spectrum peaks were produced with Lorentzian shapes, for the x-rays, and Gaussian distributions. The MGA code determined the Pu isotopes and specific power of this calculated spectrum and compared it to a similar analysis on a measured spectrum

Overview of the MCU Monte Carlo software package

International Nuclear Information System (INIS)

Kalugin, M.A.; Oleynik, D.S.; Shkarovsky, D.A.

2013-01-01

MCU (Monte Carlo Universal) is a project on development and practical use of a universal computer code for simulation of particle transport (neutrons, photons, electrons, positrons) in three-dimensional systems by means of the Monte Carlo method. This paper provides the information on the current state of the project. The developed libraries of constants are briefly described, and the potentialities of the MCU-5 package modules and the executable codes compiled from them are characterized. Examples of important problems of reactor physics solved with the code are presented. It is shown that the MCU constructor tool is able to assemble a full-scale 3D model from templates describing single components using simple and intuitive graphic user interface. The templates are prepared by a skilled user and stored in constructor's templates library. Ordinary user works with the graphic user interface and does not deal with MCU input data directly. At the present moment there are template libraries for several types of reactors

OGRE, Monte-Carlo System for Gamma Transport Problems

International Nuclear Information System (INIS)

1984-01-01

1 - Nature of physical problem solved: The OGRE programme system was designed to calculate, by Monte Carlo methods, any quantity related to gamma-ray transport. The system is represented by two examples - OGRE-P1 and OGRE-G. The OGRE-P1 programme is a simple prototype which calculates dose rate on one side of a slab due to a plane source on the other side. The OGRE-G programme, a prototype of a programme utilizing a general-geometry routine, calculates dose rate at arbitrary points. A very general source description in OGRE-G may be employed by reading a tape prepared by the user. 2 - Method of solution: Case histories of gamma rays in the prescribed geometry are generated and analyzed to produce averages of any desired quantity which, in the case of the prototypes, are gamma-ray dose rates. The system is designed to achieve generality by ease of modification. No importance sampling is built into the prototypes, a very general geometry subroutine permits the treatment of complicated geometries. This is essentially the same routine used in the O5R neutron transport system. Boundaries may be either planes or quadratic surfaces, arbitrarily oriented and intersecting in arbitrary fashion. Cross section data is prepared by the auxiliary master cross section programme XSECT which may be used to originate, update, or edit the master cross section tape. The master cross section tape is utilized in the OGRE programmes to produce detailed tables of macroscopic cross sections which are used during the Monte Carlo calculations. 3 - Restrictions on the complexity of the problem: Maximum cross-section array information may be estimated by a given formula for a specific problem. The number of regions must be less than or equal to 50

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

International Nuclear Information System (INIS)

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

2015-01-01

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

A portable, parallel, object-oriented Monte Carlo neutron transport code in C++

International Nuclear Information System (INIS)

Lee, S.R.; Cummings, J.C.; Nolen, S.D.

1997-01-01

We have developed a multi-group Monte Carlo neutron transport code using C++ and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and α-eigenvalues and is portable to and runs parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities of MC++ are discussed, along with physics and performance results on a variety of hardware, including all Accelerated Strategic Computing Initiative (ASCI) hardware. Current parallel performance indicates the ability to compute α-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed

FMCEIR: a Monte Carlo program for solving the stationary neutron and gamma transport equation

International Nuclear Information System (INIS)

Taormina, A.

1978-05-01

FMCEIR is a three-dimensional Monte Carlo program for solving the stationary neutron and gamma transport equation. It is used to study the problem of neutron and gamma streaming in the GCFR and HHT reactor channels. (G.T.H.)

(U) Introduction to Monte Carlo Methods

Energy Technology Data Exchange (ETDEWEB)

Hungerford, Aimee L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-03-20

Monte Carlo methods are very valuable for representing solutions to particle transport problems. Here we describe a “cook book” approach to handling the terms in a transport equation using Monte Carlo methods. Focus is on the mechanics of a numerical Monte Carlo code, rather than the mathematical foundations of the method.

Monte Carlo simulation of ballistic transport in high-mobility channels

Energy Technology Data Exchange (ETDEWEB)

Sabatini, G; Marinchio, H; Palermo, C; Varani, L; Daoud, T; Teissier, R [Institut d' Electronique du Sud (CNRS UMR 5214) - Universite Montpellier II (France); Rodilla, H; Gonzalez, T; Mateos, J, E-mail: sabatini@ies.univ-montp2.f [Departamento de Fisica Aplicada - Universidad de Salamanca (Spain)

2009-11-15

By means of Monte Carlo simulations coupled with a two-dimensional Poisson solver, we evaluate directly the possibility to use high mobility materials in ultra fast devices exploiting ballistic transport. To this purpose, we have calculated specific physical quantities such as the transit time, the transit velocity, the free flight time and the mean free path as functions of applied voltage in InAs channels with different lengths, from 2000 nm down to 50 nm. In this way the transition from diffusive to ballistic transport is carefully described. We remark a high value of the mean transit velocity with a maximum of 14x10{sup 5} m/s for a 50 nm-long channel and a transit time shorter than 0.1 ps, corresponding to a cutoff frequency in the terahertz domain. The percentage of ballistic electrons and the number of scatterings as functions of distance are also reported, showing the strong influence of quasi-ballistic transport in the shorter channels.

Monte Carlo simulation of ballistic transport in high-mobility channels

International Nuclear Information System (INIS)

Sabatini, G; Marinchio, H; Palermo, C; Varani, L; Daoud, T; Teissier, R; Rodilla, H; Gonzalez, T; Mateos, J

2009-01-01

By means of Monte Carlo simulations coupled with a two-dimensional Poisson solver, we evaluate directly the possibility to use high mobility materials in ultra fast devices exploiting ballistic transport. To this purpose, we have calculated specific physical quantities such as the transit time, the transit velocity, the free flight time and the mean free path as functions of applied voltage in InAs channels with different lengths, from 2000 nm down to 50 nm. In this way the transition from diffusive to ballistic transport is carefully described. We remark a high value of the mean transit velocity with a maximum of 14x10 5 m/s for a 50 nm-long channel and a transit time shorter than 0.1 ps, corresponding to a cutoff frequency in the terahertz domain. The percentage of ballistic electrons and the number of scatterings as functions of distance are also reported, showing the strong influence of quasi-ballistic transport in the shorter channels.

Coupled-resonator waveguide perfect transport single-photon by interatomic dipole-dipole interaction

Science.gov (United States)

Yan, Guo-an; Lu, Hua; Qiao, Hao-xue; Chen, Ai-xi; Wu, Wan-qing

2018-06-01

We theoretically investigate single-photon coherent transport in a one-dimensional coupled-resonator waveguide coupled to two quantum emitters with dipole-dipole interactions. The numerical simulations demonstrate that the transmission spectrum of the photon depends on the two atoms dipole-dipole interactions and the photon-atom couplings. The dipole-dipole interactions may change the dip positions in the spectra and the coupling strength may broaden the frequency band width in the transmission spectrum. We further demonstrate that the typical transmission spectra split into two dips due to the dipole-dipole interactions. This phenomenon may be used to manufacture new quantum waveguide devices.

BOMAB phantom manufacturing quality assurance study using Monte Carlo computations

International Nuclear Information System (INIS)

Mallett, M.W.

1994-01-01

Monte Carlo calculations have been performed to assess the importance of and quantify quality assurance protocols in the manufacturing of the Bottle-Manikin-Absorption (BOMAB) phantom for calibrating in vivo measurement systems. The parameters characterizing the BOMAB phantom that were examined included height, fill volume, fill material density, wall thickness, and source concentration. Transport simulation was performed for monoenergetic photon sources of 0.200, 0.662, and 1,460 MeV. A linear response was observed in the photon current exiting the exterior surface of the BOMAB phantom due to variations in these parameters. Sensitivity studies were also performed for an in vivo system in operation at the Pacific Northwest Laboratories in Richland, WA. Variations in detector current for this in vivo system are reported for changes in the BOMAB phantom parameters studied here. Physical justifications for the observed results are also discussed

Acceleration of Monte Carlo simulation of photon migration in complex heterogeneous media using Intel many-integrated core architecture.

Science.gov (United States)

Gorshkov, Anton V; Kirillin, Mikhail Yu

2015-08-01

Over two decades, the Monte Carlo technique has become a gold standard in simulation of light propagation in turbid media, including biotissues. Technological solutions provide further advances of this technique. The Intel Xeon Phi coprocessor is a new type of accelerator for highly parallel general purpose computing, which allows execution of a wide range of applications without substantial code modification. We present a technical approach of porting our previously developed Monte Carlo (MC) code for simulation of light transport in tissues to the Intel Xeon Phi coprocessor. We show that employing the accelerator allows reducing computational time of MC simulation and obtaining simulation speed-up comparable to GPU. We demonstrate the performance of the developed code for simulation of light transport in the human head and determination of the measurement volume in near-infrared spectroscopy brain sensing.

COG10, Multiparticle Monte Carlo Code System for Shielding and Criticality Use

International Nuclear Information System (INIS)

2007-01-01

1 - Description of program or function: COG is a modern, full-featured Monte Carlo radiation transport code which provides accurate answers to complex shielding, criticality, and activation problems. COG was written to be state-of-the-art and free of physics approximations and compromises found in earlier codes. COG is fully 3-D, uses point-wise cross sections and exact angular scattering, and allows a full range of biasing options to speed up solutions for deep penetration problems. Additionally, a criticality option is available for computing Keff for assemblies of fissile materials. ENDL or ENDFB cross section libraries may be used. COG home page: http://www-phys.llnl.gov/N_Div/COG/. Cross section libraries are included in the package. COG can use either the LLNL ENDL-90 cross section set or the ENDFB/VI set. Analytic surfaces are used to describe geometric boundaries. Parts (volumes) are described by a method of Constructive Solid Geometry. Surface types include surfaces of up to fourth order, and pseudo-surfaces such as boxes, finite cylinders, and figures of revolution. Repeated assemblies need be defined only once. Parts are visualized in cross-section and perspective picture views. Source and random-walk biasing techniques may be selected to improve solution statistics. These include source angular biasing, importance weighting, particle splitting and Russian roulette, path-length stretching, point detectors, scattered direction biasing, and forced collisions. Criticality - For a fissioning system, COG will compute Keff by transporting batches of neutrons through the system. Activation - COG can compute gamma-ray doses due to neutron-activated materials, starting with just a neutron source. Coupled Problems - COG can solve coupled problems involving neutrons, photons, and electrons. 2 - Methods:COG uses Monte Carlo methods to solve the Boltzmann transport equation for particles traveling through arbitrary 3-dimensional geometries. Neutrons, photons

Forest canopy BRDF simulation using Monte Carlo method

NARCIS (Netherlands)

Huang, J.; Wu, B.; Zeng, Y.; Tian, Y.

2006-01-01

Monte Carlo method is a random statistic method, which has been widely used to simulate the Bidirectional Reflectance Distribution Function (BRDF) of vegetation canopy in the field of visible remote sensing. The random process between photons and forest canopy was designed using Monte Carlo method.

A punctual flux estimator and reactions rates optimization in neutral particles transport calculus by the Monte Carlo method

International Nuclear Information System (INIS)

Authier, N.

1998-12-01

One of the questions asked in radiation shielding problems is the estimation of the radiation level in particular to determine accessibility of working persons in controlled area (nuclear power plants, nuclear fuel reprocessing plants) or to study the dose gradients encountered in material (iron nuclear vessel, medical therapy, electronics in satellite). The flux and reaction rate estimators used in Monte Carlo codes give average values in volumes or on surfaces of the geometrical description of the system. But in certain configurations, the knowledge of punctual deposited energy and dose estimates are necessary. The Monte Carlo estimate of the flux at a point of interest is a calculus which presents an unbounded variance. The central limit theorem cannot be applied thus no easy confidence level may be calculated. The convergence rate is then very poor. We propose in this study a new solution for the photon flux at a point estimator. The method is based on the 'once more collided flux estimator' developed earlier for neutron calculations. It solves the problem of the unbounded variance and do not add any bias to the estimation. We show however that our new sampling schemes specially developed to treat the anisotropy of the photon coherent scattering is necessary for a good and regular behavior of the estimator. This developments integrated in the TRIPOLI-4 Monte Carlo code add the possibility of an unbiased punctual estimate on media interfaces. (author)

Asymptotic equilibrium diffusion analysis of time-dependent Monte Carlo methods for grey radiative transfer

International Nuclear Information System (INIS)

Densmore, Jeffery D.; Larsen, Edward W.

2004-01-01

The equations of nonlinear, time-dependent radiative transfer are known to yield the equilibrium diffusion equation as the leading-order solution of an asymptotic analysis when the mean-free path and mean-free time of a photon become small. We apply this same analysis to the Fleck-Cummings, Carter-Forest, and N'kaoua Monte Carlo approximations for grey (frequency-independent) radiative transfer. Although Monte Carlo simulation usually does not require the discretizations found in deterministic transport techniques, Monte Carlo methods for radiative transfer require a time discretization due to the nonlinearities of the problem. If an asymptotic analysis of the equations used by a particular Monte Carlo method yields an accurate time-discretized version of the equilibrium diffusion equation, the method should generate accurate solutions if a time discretization is chosen that resolves temperature changes, even if the time steps are much larger than the mean-free time of a photon. This analysis is of interest because in many radiative transfer problems, it is a practical necessity to use time steps that are large compared to a mean-free time. Our asymptotic analysis shows that: (i) the N'kaoua method has the equilibrium diffusion limit, (ii) the Carter-Forest method has the equilibrium diffusion limit if the material temperature change during a time step is small, and (iii) the Fleck-Cummings method does not have the equilibrium diffusion limit. We include numerical results that verify our theoretical predictions

Monte-Carlo simulation of complex vapor-transport systems for RIB applications

International Nuclear Information System (INIS)

Zhang, Y.; Alton, G.D.

2005-01-01

In order to minimize decay losses of short-lived radioactive species at ISOL based RIB facilities, effusive-flow particle transit times between target and ion source must be as short as practically achievable. A Monte-Carlo code has been developed for simulating the effusive-flow of neutral particles through vapor-transport systems independent of materials of construction. The code provides average distance traveled and time information associated with the transit of individual particles through a system. It offers a cost effective and accurate means for arriving at vapor-transport system designs. In this report, the code will be described and results obtained by its use in evaluating several prototype vapor-transport systems using specular reflection, cosine and isotropic particle re-emission about the normal to the surface models following adsorption. Simulation results obtained with an isotropic distribution are in close agreement with experimental measurements of the properties of prototype vapor-transport systems fabricated at the Holifield Radioactive Ion Beam Facility

A single-source photon source model of a linear accelerator for Monte Carlo dose calculation.

Science.gov (United States)

Nwankwo, Obioma; Glatting, Gerhard; Wenz, Frederik; Fleckenstein, Jens

2017-01-01

To introduce a new method of deriving a virtual source model (VSM) of a linear accelerator photon beam from a phase space file (PSF) for Monte Carlo (MC) dose calculation. A PSF of a 6 MV photon beam was generated by simulating the interactions of primary electrons with the relevant geometries of a Synergy linear accelerator (Elekta AB, Stockholm, Sweden) and recording the particles that reach a plane 16 cm downstream the electron source. Probability distribution functions (PDFs) for particle positions and energies were derived from the analysis of the PSF. These PDFs were implemented in the VSM using inverse transform sampling. To model particle directions, the phase space plane was divided into a regular square grid. Each element of the grid corresponds to an area of 1 mm2 in the phase space plane. The average direction cosines, Pearson correlation coefficient (PCC) between photon energies and their direction cosines, as well as the PCC between the direction cosines were calculated for each grid element. Weighted polynomial surfaces were then fitted to these 2D data. The weights are used to correct for heteroscedasticity across the phase space bins. The directions of the particles created by the VSM were calculated from these fitted functions. The VSM was validated against the PSF by comparing the doses calculated by the two methods for different square field sizes. The comparisons were performed with profile and gamma analyses. The doses calculated with the PSF and VSM agree to within 3% /1 mm (>95% pixel pass rate) for the evaluated fields. A new method of deriving a virtual photon source model of a linear accelerator from a PSF file for MC dose calculation was developed. Validation results show that the doses calculated with the VSM and the PSF agree to within 3% /1 mm.

A single-source photon source model of a linear accelerator for Monte Carlo dose calculation.

Directory of Open Access Journals (Sweden)

Obioma Nwankwo

Full Text Available To introduce a new method of deriving a virtual source model (VSM of a linear accelerator photon beam from a phase space file (PSF for Monte Carlo (MC dose calculation.A PSF of a 6 MV photon beam was generated by simulating the interactions of primary electrons with the relevant geometries of a Synergy linear accelerator (Elekta AB, Stockholm, Sweden and recording the particles that reach a plane 16 cm downstream the electron source. Probability distribution functions (PDFs for particle positions and energies were derived from the analysis of the PSF. These PDFs were implemented in the VSM using inverse transform sampling. To model particle directions, the phase space plane was divided into a regular square grid. Each element of the grid corresponds to an area of 1 mm2 in the phase space plane. The average direction cosines, Pearson correlation coefficient (PCC between photon energies and their direction cosines, as well as the PCC between the direction cosines were calculated for each grid element. Weighted polynomial surfaces were then fitted to these 2D data. The weights are used to correct for heteroscedasticity across the phase space bins. The directions of the particles created by the VSM were calculated from these fitted functions. The VSM was validated against the PSF by comparing the doses calculated by the two methods for different square field sizes. The comparisons were performed with profile and gamma analyses.The doses calculated with the PSF and VSM agree to within 3% /1 mm (>95% pixel pass rate for the evaluated fields.A new method of deriving a virtual photon source model of a linear accelerator from a PSF file for MC dose calculation was developed. Validation results show that the doses calculated with the VSM and the PSF agree to within 3% /1 mm.

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

DEFF Research Database (Denmark)

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

2015-01-01

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

Investigation of pattern recognition techniques for the indentification of splitting surfaces in Monte Carlo particle transport calculations

International Nuclear Information System (INIS)

Macdonald, J.L.

1975-08-01

Statistical and deterministic pattern recognition systems are designed to classify the state space of a Monte Carlo transport problem into importance regions. The surfaces separating the regions can be used for particle splitting and Russian roulette in state space in order to reduce the variance of the Monte Carlo tally. Computer experiments are performed to evaluate the performance of the technique using one and two dimensional Monte Carlo problems. Additional experiments are performed to determine the sensitivity of the technique to various pattern recognition and Monte Carlo problem dependent parameters. A system for applying the technique to a general purpose Monte Carlo code is described. An estimate of the computer time required by the technique is made in order to determine its effectiveness as a variance reduction device. It is recommended that the technique be further investigated in a general purpose Monte Carlo code. (auth)

Analysis of QUADOS problem on TLD-ALBEDO personal dosemeter responses using discrete ordinates and Monte Carlo methods

International Nuclear Information System (INIS)

Kodeli, I.; Tanner, R.

2005-01-01

In the scope of QUADOS, a Concerted Action of the European Commission, eight calculational problems were prepared in order to evaluate the use of computational codes for dosimetry in radiation protection and medical physics, and to disseminate 'good practice' throughout the radiation dosimetry community. This paper focuses on the analysis of the P4 problem on the 'TLD-albedo dosemeter: neutron and/or photon response of a four-element TL-dosemeter mounted on a standard ISO slab phantom'. Altogether 17 solutions were received from the participants, 14 of those transported neutrons and 15 photons. Most participants (16 out of 17) used Monte Carlo methods. These calculations are time-consuming, requiring several days of CPU time to perform the whole set of calculations and achieve good statistical precision. The possibility of using deterministic discrete ordinates codes as an alternative to Monte Carlo was therefore investigated and is presented here. In particular the capacity of the adjoint mode calculations is shown. (authors)

The electron transport problem sampling by Monte Carlo individual collision technique

International Nuclear Information System (INIS)

Androsenko, P.A.; Belousov, V.I.

2005-01-01

The problem of electron transport is of most interest in all fields of the modern science. To solve this problem the Monte Carlo sampling has to be used. The electron transport is characterized by a large number of individual interactions. To simulate electron transport the 'condensed history' technique may be used where a large number of collisions are grouped into a single step to be sampled randomly. Another kind of Monte Carlo sampling is the individual collision technique. In comparison with condensed history technique researcher has the incontestable advantages. For example one does not need to give parameters altered by condensed history technique like upper limit for electron energy, resolution, number of sub-steps etc. Also the condensed history technique may lose some very important tracks of electrons because of its limited nature by step parameters of particle movement and due to weakness of algorithms for example energy indexing algorithm. There are no these disadvantages in the individual collision technique. This report presents some sampling algorithms of new version BRAND code where above mentioned technique is used. All information on electrons was taken from Endf-6 files. They are the important part of BRAND. These files have not been processed but directly taken from electron information source. Four kinds of interaction like the elastic interaction, the Bremsstrahlung, the atomic excitation and the atomic electro-ionization were considered. In this report some results of sampling are presented after comparison with analogs. For example the endovascular radiotherapy problem (P2) of QUADOS2002 was presented in comparison with another techniques that are usually used. (authors)

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.

Thermal transport in nanocrystalline Si and SiGe by ab initio based Monte Carlo simulation.

Science.gov (United States)

Yang, Lina; Minnich, Austin J

2017-03-14

Nanocrystalline thermoelectric materials based on Si have long been of interest because Si is earth-abundant, inexpensive, and non-toxic. However, a poor understanding of phonon grain boundary scattering and its effect on thermal conductivity has impeded efforts to improve the thermoelectric figure of merit. Here, we report an ab-initio based computational study of thermal transport in nanocrystalline Si-based materials using a variance-reduced Monte Carlo method with the full phonon dispersion and intrinsic lifetimes from first-principles as input. By fitting the transmission profile of grain boundaries, we obtain excellent agreement with experimental thermal conductivity of nanocrystalline Si [Wang et al. Nano Letters 11, 2206 (2011)]. Based on these calculations, we examine phonon transport in nanocrystalline SiGe alloys with ab-initio electron-phonon scattering rates. Our calculations show that low energy phonons still transport substantial amounts of heat in these materials, despite scattering by electron-phonon interactions, due to the high transmission of phonons at grain boundaries, and thus improvements in ZT are still possible by disrupting these modes. This work demonstrates the important insights into phonon transport that can be obtained using ab-initio based Monte Carlo simulations in complex nanostructured materials.

Robust Adaptive Photon Tracing using Photon Path Visibility

DEFF Research Database (Denmark)

Hachisuka, Toshiya; Jensen, Henrik Wann

2011-01-01

We present a new adaptive photon tracing algorithm which can handle illumination settings that are considered difficult for photon tracing approaches such as outdoor scenes, close-ups of a small part of an illuminated region, and illumination coming through a small gap. The key contribution in our...... algorithm is the use of visibility of photon path as the importance function which ensures that our sampling algorithm focuses on paths that are visible from the given viewpoint. Our sampling algorithm builds on two recent developments in Markov chain Monte Carlo methods: adaptive Markov chain sampling...... and replica exchange. Using these techniques, each photon path is adaptively mutated and it explores the sampling space efficiently without being stuck at a local peak of the importance function. We have implemented this sampling approach in the progressive photon mapping algorithm which provides visibility...

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

Directory of Open Access Journals (Sweden)

Ilić Radovan D.

2002-01-01

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

Application of artificial intelligence techniques to the acceleration of Monte Carlo transport calculations

International Nuclear Information System (INIS)

Maconald, J.L.; Cashwell, E.D.

1978-09-01

The techniques of learning theory and pattern recognition are used to learn splitting surface locations for the Monte Carlo neutron transport code MCN. A study is performed to determine default values for several pattern recognition and learning parameters. The modified MCN code is used to reduce computer cost for several nontrivial example problems

Monte Carlo simulation of radioactive contaminant transport in unsaturated porous media

International Nuclear Information System (INIS)

Giacobbo, F.; Patelli, E.; Zio, E.

2005-01-01

In the current proposed solutions of radioactive waste repositories, the protective function against the radionuclide water-driven transport back to the biosphere is to be provided by an integrated system of artificial and natural geologic barriers. The complexity of the transport process in the barriers' heterogeneous media forces approximations to the classical analytical-numerical models, thus reducing their adherence to reality. In an attempt to overcome these difficulties, in the present paper we adopt a Monte Carlo simulation approach, previously developed on the basis of the Kolmogorov and Dmitriev theory of branching stochastic processes. The approach is here extended for describing transport through unsaturated porous media under unsteady flow conditions. This generalization entails the determination of the functional dependence of the parameters of the proposed transport model from the water content, which changes in space and time during the water infiltration process. The approach is verified with respect to a case of non-reactive transport under transient unsaturated field conditions by a comparison with a standard code based on the classical advection-dispersion equations. An application regarding linear reactive transport is then presented. (authors)

Monte Carlo Modeling of Dual and Triple Photon Energy Absorptiometry Technique

Directory of Open Access Journals (Sweden)

Alireza Kamali-Asl

2007-12-01

Full Text Available Introduction: Osteoporosis is a bone disease in which there is a reduction in the amount of bone mineral content leading to an increase in the risk of bone fractures. The affected individuals not only have to go through lots of pain and suffering but this disease also results in high economic costs to the society due to a large number of fractures.  A timely and accurate diagnosis of this disease makes it possible to start a treatment and thus preventing bone fractures as a result of osteoporosis. Radiographic methods are particularly well suited for in vivo determination of bone mineral density (BMD due to the relatively high x-ray absorption properties of bone mineral compared to other tissues. Materials and Methods: Monte Carlo simulation has been conducted to explore the possibilities of triple photon energy absorptiometry (TPA in the measurement of bone mineral content. The purpose of this technique is to correctly measure the bone mineral density in the presence of fatty and soft tissues. The same simulations have been done for a dual photon energy absorptiometry (DPA system and an extended DPA system. Results: Using DPA with three components improves the accuracy of the obtained result while the simulation results show that TPA system is not accurate enough to be considered as an adequate method for the measurement of bone mineral density. Discussion: The reason for the improvement in the accuracy is the consideration of fatty tissue in TPA method while having attenuation coefficient as a function of energy makes TPA an inadequate method. Conclusion: Using TPA method is not a perfect solution to overcome the problem of non uniformity in the distribution of fatty tissue.

MCNP4C2, Coupled Neutron, Electron Gamma 3-D Time-Dependent Monte Carlo Transport Calculations

International Nuclear Information System (INIS)

2002-01-01

1 - Description of program or function: MCNP is a general-purpose, continuous-energy, generalized geometry, time-dependent, coupled neutron-photon-electron Monte Carlo transport code system. MCNP4C2 is an interim release of MCNP4C with distribution restricted to the Criticality Safety community and attendees of the LANL MCNP workshops. The major new features of MCNP4C2 include: - Photonuclear physics; - Interactive plotting; - Plot superimposed weight window mesh; - Implement remaining macro-body surfaces; - Upgrade macro-bodies to surface sources and other capabilities; - Revised summary tables; - Weight window improvements. See the MCNP home page more information http://www-xdiv.lanl.gov/XCI/PROJECTS/MCNP with a link to the MCNP Forum. See the Electronic Notebook at http://www-rsicc.ornl.gov/rsic.html for information on user experiences with MCNP. 2 - Methods:MCNP treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces. Pointwise continuous-energy cross section data are used, although multigroup data may also be used. Fixed-source adjoint calculations may be made with the multigroup data option. For neutrons, all reactions in a particular cross-section evaluation are accounted for. Both free gas and S(alpha, beta) thermal treatments are used. Criticality sources as well as fixed and surface sources are available. For photons, the code takes account of incoherent and coherent scattering with and without electron binding effects, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. A very general source and tally structure is available. The tallies have extensive statistical analysis of convergence. Rapid convergence is enabled by a wide variety of variance reduction methods. Energy ranges are 0-60 MeV for neutrons (data generally only available up to

Monte Carlo investigation of anomalous transport in presence of a discontinuity and of an advection field

Science.gov (United States)

Marseguerra, M.; Zoia, A.

2007-04-01

Anomalous diffusion has recently turned out to be almost ubiquitous in transport problems. When the physical properties of the medium where the transport process takes place are stationary and constant at each spatial location, anomalous transport has been successfully analysed within the Continuous Time Random Walk (CTRW) model. In this paper, within a Monte Carlo approach to CTRW, we focus on the particle transport through two regions characterized by different physical properties, in presence of an external driving action constituted by an additional advective field, modelled within both the Galilei invariant and Galilei variant schemes. Particular attention is paid to the interplay between the distributions of space and time across the discontinuity. The resident concentration and the flux of the particles are finally evaluated and it is shown that at the interface between the two regions the flux is continuous as required by mass conservation, while the concentration may reveal a neat discontinuity. This result could open the route to the Monte Carlo investigation of the effectiveness of a physical discontinuity acting as a filter on particle concentration.

Evaluation of Monte Carlo electron-Transport algorithms in the integrated Tiger series codes for stochastic-media simulations

International Nuclear Information System (INIS)

Franke, B.C.; Kensek, R.P.; Prinja, A.K.

2013-01-01

Stochastic-media simulations require numerous boundary crossings. We consider two Monte Carlo electron transport approaches and evaluate accuracy with numerous material boundaries. In the condensed-history method, approximations are made based on infinite-medium solutions for multiple scattering over some track length. Typically, further approximations are employed for material-boundary crossings where infinite-medium solutions become invalid. We have previously explored an alternative 'condensed transport' formulation, a Generalized Boltzmann-Fokker-Planck (GBFP) method, which requires no special boundary treatment but instead uses approximations to the electron-scattering cross sections. Some limited capabilities for analog transport and a GBFP method have been implemented in the Integrated Tiger Series (ITS) codes. Improvements have been made to the condensed history algorithm. The performance of the ITS condensed-history and condensed-transport algorithms are assessed for material-boundary crossings. These assessments are made both by introducing artificial material boundaries and by comparison to analog Monte Carlo simulations. (authors)

Acceleration of a Monte Carlo radiation transport code

International Nuclear Information System (INIS)

Hochstedler, R.D.; Smith, L.M.

1996-01-01

Execution time for the Integrated TIGER Series (ITS) Monte Carlo radiation transport code has been reduced by careful re-coding of computationally intensive subroutines. Three test cases for the TIGER (1-D slab geometry), CYLTRAN (2-D cylindrical geometry), and ACCEPT (3-D arbitrary geometry) codes were identified and used to benchmark and profile program execution. Based upon these results, sixteen top time-consuming subroutines were examined and nine of them modified to accelerate computations with equivalent numerical output to the original. The results obtained via this study indicate that speedup factors of 1.90 for the TIGER code, 1.67 for the CYLTRAN code, and 1.11 for the ACCEPT code are achievable. copyright 1996 American Institute of Physics

Quantitative analysis of optical properties of flowing blood using a photon-cell interactive Monte Carlo code: effects of red blood cells' orientation on light scattering.

Science.gov (United States)

Sakota, Daisuke; Takatani, Setsuo

2012-05-01

Optical properties of flowing blood were analyzed using a photon-cell interactive Monte Carlo (pciMC) model with the physical properties of the flowing red blood cells (RBCs) such as cell size, shape, refractive index, distribution, and orientation as the parameters. The scattering of light by flowing blood at the He-Ne laser wavelength of 632.8 nm was significantly affected by the shear rate. The light was scattered more in the direction of flow as the flow rate increased. Therefore, the light intensity transmitted forward in the direction perpendicular to flow axis decreased. The pciMC model can duplicate the changes in the photon propagation due to moving RBCs with various orientations. The resulting RBC's orientation that best simulated the experimental results was with their long axis perpendicular to the direction of blood flow. Moreover, the scattering probability was dependent on the orientation of the RBCs. Finally, the pciMC code was used to predict the hematocrit of flowing blood with accuracy of approximately 1.0 HCT%. The photon-cell interactive Monte Carlo (pciMC) model can provide optical properties of flowing blood and will facilitate the development of the non-invasive monitoring of blood in extra corporeal circulatory systems.

Improving the Ar I and II branching ratio calibration method: Monte Carlo simulations of effects from photon scattering/reflecting in hollow cathodes

Science.gov (United States)

Lawler, J. E.; Den Hartog, E. A.

2018-03-01

The Ar I and II branching ratio calibration method is discussed with the goal of improving the technique. This method of establishing a relative radiometric calibration is important in ongoing research to improve atomic transition probabilities for quantitative spectroscopy in astrophysics and other fields. Specific suggestions are presented along with Monte Carlo simulations of wavelength dependent effects from scattering/reflecting of photons in a hollow cathode.

EGS code system: computer programs for the Monte Carlo simulation of electromagnetic cascade showers. Version 3

International Nuclear Information System (INIS)

Ford, R.L.; Nelson, W.R.

1978-06-01

A code to simulate almost any electron--photon transport problem conceivable is described. The report begins with a lengthy historical introduction and a description of the shower generation process. Then the detailed physics of the shower processes and the methods used to simulate them are presented. Ideas of sampling theory, transport techniques, particle interactions in general, and programing details are discussed. Next, EGS calculations and various experiments and other Monte Carlo results are compared. The remainder of the report consists of user manuals for EGS, PEGS, and TESTSR codes; options, input specifications, and typical output are included. 38 figures, 12 tables

Monte Carlo investigation of minority electron transport in InP

International Nuclear Information System (INIS)

Osman, M.A.; Grubin, H.L.

1989-01-01

This paper discusses the investigation of the transport of minority electrons in p-type InP for acceptor doping level of 10 18 cm 3 using Monte Carlo procedures. It is found that the velocity of minority electrons are significantly lower than that of majority electrons for fields below 15 kV/cm and slightly higher at higher fields. The study shows that the interaction between the electrons and majority holes leads to reducing the mobility of electrons from 2000 cm 2 /Vs to 1500 cm 2 /Vs

Reliability analysis of neutron transport simulation using Monte Carlo method

International Nuclear Information System (INIS)

Souza, Bismarck A. de; Borges, Jose C.

1995-01-01

This work presents a statistical and reliability analysis covering data obtained by computer simulation of neutron transport process, using the Monte Carlo method. A general description of the method and its applications is presented. Several simulations, corresponding to slowing down and shielding problems have been accomplished. The influence of the physical dimensions of the materials and of the sample size on the reliability level of results was investigated. The objective was to optimize the sample size, in order to obtain reliable results, optimizing computation time. (author). 5 refs, 8 figs

Monte Carlo studies on photon interactions in radiobiological experiments

Science.gov (United States)

Shahmohammadi Beni, Mehrdad; Krstic, D.; Nikezic, D.

2018-01-01

X-ray and γ-ray photons have been widely used for studying radiobiological effects of ionizing radiations. Photons are indirectly ionizing radiations so they need to set in motion electrons (which are a directly ionizing radiation) to perform the ionizations. When the photon dose decreases to below a certain limit, the number of electrons set in motion will become so small that not all cells in an “exposed” cell population can get at least one electron hit. When some cells in a cell population are not hit by a directly ionizing radiation (in other words not irradiated), there will be rescue effect between the irradiated cells and non-irradiated cells, and the resultant radiobiological effect observed for the “exposed” cell population will be different. In the present paper, the mechanisms underlying photon interactions in radiobiological experiments were studied using our developed NRUphoton computer code, which was benchmarked against the MCNP5 code by comparing the photon dose delivered to the cell layer underneath the water medium. The following conclusions were reached: (1) The interaction fractions decreased in the following order: 16O > 12C > 14N > 1H. Bulges in the interaction fractions (versus water medium thickness) were observed, which reflected changes in the energies of the propagating photons due to traversals of different amount of water medium as well as changes in the energy-dependent photon interaction cross-sections. (2) Photoelectric interaction and incoherent scattering dominated for lower-energy (10 keV) and high-energy (100 keV and 1 MeV) incident photons. (3) The fractions of electron ejection from different nuclei were mainly governed by the photoelectric effect cross-sections, and the fractions from the 1s subshell were the largest. (4) The penetration fractions in general decreased with increasing medium thickness, and increased with increasing incident photon energy, the latter being explained by the corresponding reduction in

Monte Carlo studies on photon interactions in radiobiological experiments.

Directory of Open Access Journals (Sweden)

Mehrdad Shahmohammadi Beni

Full Text Available X-ray and γ-ray photons have been widely used for studying radiobiological effects of ionizing radiations. Photons are indirectly ionizing radiations so they need to set in motion electrons (which are a directly ionizing radiation to perform the ionizations. When the photon dose decreases to below a certain limit, the number of electrons set in motion will become so small that not all cells in an "exposed" cell population can get at least one electron hit. When some cells in a cell population are not hit by a directly ionizing radiation (in other words not irradiated, there will be rescue effect between the irradiated cells and non-irradiated cells, and the resultant radiobiological effect observed for the "exposed" cell population will be different. In the present paper, the mechanisms underlying photon interactions in radiobiological experiments were studied using our developed NRUphoton computer code, which was benchmarked against the MCNP5 code by comparing the photon dose delivered to the cell layer underneath the water medium. The following conclusions were reached: (1 The interaction fractions decreased in the following order: 16O > 12C > 14N > 1H. Bulges in the interaction fractions (versus water medium thickness were observed, which reflected changes in the energies of the propagating photons due to traversals of different amount of water medium as well as changes in the energy-dependent photon interaction cross-sections. (2 Photoelectric interaction and incoherent scattering dominated for lower-energy (10 keV and high-energy (100 keV and 1 MeV incident photons. (3 The fractions of electron ejection from different nuclei were mainly governed by the photoelectric effect cross-sections, and the fractions from the 1s subshell were the largest. (4 The penetration fractions in general decreased with increasing medium thickness, and increased with increasing incident photon energy, the latter being explained by the corresponding reduction in

A kinematic fit method for all-photon events

International Nuclear Information System (INIS)

Du Shuxian; Yuan Changzheng; Chinese Academy of Sciences, Beijing

2006-01-01

An improved kinematic fit method is developed for analyzing all-photon events, where the interaction point is unknown. The fitting algorithm is checked with Monte Carlo samples to ensure that the fitting program works properly. This is applied to the Monte Carlo simulated ψ(2S) decays. A higher efficiency is achieved. This method can be generally applied to analyzing all-photon events at electron-positron collider. (authors)

Correction factors for photon beam quality for cylindrical ionization chambers: Monte Carlo calculations by using the PENELOPE code

International Nuclear Information System (INIS)

Barreras Caballero, A. A.; Hernandez Garcia, J.J.; Alfonso Laguardia, R.

2009-01-01

Were directly determined correction factors depending on the type camera beam quality, k, Q, and kQ, Qo, instead of the product (w, air p) Q, for three type cylindrical ionization chambers Pinpoint and divergent monoenergetic beams of photons in a wide range of energies (4-20 MV). The method of calculation used dispenses with the approaches taken in the classic procedure considered independent of braking power ratios and the factors disturbance of the camera. A detailed description of the geometry and materials chambers were supplied by the manufacturer and used as data input for the system 2006 of PENELOPE Monte Carlo calculation using a User code that includes correlated sampling, and forced interactions division of particles. We used a photon beam Co-60 as beam reference for calculating the correction factors for beam quality. No data exist for the cameras PTW 31014, 31015 and 31016 in the TRS-398 at they do not compare the results with data calculated or determined experimentally by other authors. (author)

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

Photon beam convolution using polyenergetic energy deposition kernels

International Nuclear Information System (INIS)

Hoban, P.W.; Murray, D.C.; Round, W.H.

1994-01-01

In photon beam convolution calculations where polyenergetic energy deposition kernels (EDKs) are used, the primary photon energy spectrum should be correctly accounted for in Monte Carlo generation of EDKs. This requires the probability of interaction, determined by the linear attenuation coefficient, μ, to be taken into account when primary photon interactions are forced to occur at the EDK origin. The use of primary and scattered EDKs generated with a fixed photon spectrum can give rise to an error in the dose calculation due to neglecting the effects of beam hardening with depth. The proportion of primary photon energy that is transferred to secondary electrons increases with depth of interaction, due to the increase in the ratio μ ab /μ as the beam hardens. Convolution depth-dose curves calculated using polyenergetic EDKs generated for the primary photon spectra which exist at depths of 0, 20 and 40 cm in water, show a fall-off which is too steep when compared with EGS4 Monte Carlo results. A beam hardening correction factor applied to primary and scattered 0 cm EDKs, based on the ratio of kerma to terma at each depth, gives primary, scattered and total dose in good agreement with Monte Carlo results. (Author)

Monte Carlo simulations of the particle transport in semiconductor detectors of fast neutrons

International Nuclear Information System (INIS)

Sedlačková, Katarína; Zaťko, Bohumír; Šagátová, Andrea; Nečas, Vladimír

2013-01-01

Several Monte Carlo all-particle transport codes are under active development around the world. In this paper we focused on the capabilities of the MCNPX code (Monte Carlo N-Particle eXtended) to follow the particle transport in semiconductor detector of fast neutrons. Semiconductor detector based on semi-insulating GaAs was the object of our investigation. As converter material capable to produce charged particles from the (n, p) interaction, a high-density polyethylene (HDPE) was employed. As the source of fast neutrons, the 239 Pu–Be neutron source was used in the model. The simulations were performed using the MCNPX code which makes possible to track not only neutrons but also recoiled protons at all interesting energies. Hence, the MCNPX code enables seamless particle transport and no other computer program is needed to process the particle transport. The determination of the optimal thickness of the conversion layer and the minimum thickness of the active region of semiconductor detector as well as the energy spectra simulation were the principal goals of the computer modeling. Theoretical detector responses showed that the best detection efficiency can be achieved for 500 μm thick HDPE converter layer. The minimum detector active region thickness has been estimated to be about 400 μm. -- Highlights: ► Application of the MCNPX code for fast neutron detector design is demonstrated. ► Simulations of the particle transport through conversion film of HDPE are presented. ► Simulations of the particle transport through detector active region are presented. ► The optimal thickness of the HDPE conversion film has been calculated. ► Detection efficiency of 0.135% was reached for 500 μm thick HDPE conversion film

Monte Carlo simulation of radiation treatment machine heads

International Nuclear Information System (INIS)

Mohan, R.

1988-01-01

Monte Carlo simulations of radiation treatment machine heads provide practical means for obtaining energy spectra and angular distributions of photons and electrons. So far, most of the work published in the literature has been limited to photons and the contaminant electrons knocked out by photons. This chapter will be confined to megavoltage photon beams produced by medical linear accelerators and 60 Co teletherapy units. The knowledge of energy spectra and angular distributions of photons and contaminant electrons emerging from such machines is important for a variety of applications in radiation dosimetry

Applications guide to the RSIC-distributed version of the MCNP code (coupled Monte Carlo neutron-photon Code)

International Nuclear Information System (INIS)

Cramer, S.N.

1985-09-01

An overview of the RSIC-distributed version of the MCNP code (a soupled Monte Carlo neutron-photon code) is presented. All general features of the code, from machine hardware requirements to theoretical details, are discussed. The current nuclide cross-section and other libraries available in the standard code package are specified, and a realistic example of the flexible geometry input is given. Standard and nonstandard source, estimator, and variance-reduction procedures are outlined. Examples of correct usage and possible misuse of certain code features are presented graphically and in standard output listings. Finally, itemized summaries of sample problems, various MCNP code documentation, and future work are given

The calculation of dose from external photon exposures using reference human phantoms and Monte Carlo methods. Pt. 7. Organ doses due to parallel and environmental exposure geometries

Energy Technology Data Exchange (ETDEWEB)

Zankl, M. [GSF - Forschungszentrum fuer Umwelt und Gesundheit Neuherberg GmbH, Oberschleissheim (Germany). Inst. fuer Strahlenschutz; Drexler, G. [GSF - Forschungszentrum fuer Umwelt und Gesundheit Neuherberg GmbH, Oberschleissheim (Germany). Inst. fuer Strahlenschutz; Petoussi-Henss, N. [GSF - Forschungszentrum fuer Umwelt und Gesundheit Neuherberg GmbH, Oberschleissheim (Germany). Inst. fuer Strahlenschutz; Saito, K. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

1997-03-01

This report presents a tabulation of organ and tissue equivalent dose as well as effective dose conversion coefficients, normalised to air kerma free in air, for occupational exposures and environmental exposures of the public to external photon radiation. For occupational exposures, whole-body irradiation with idealised geometries, i.e. broad parallel beams and fully isotropic radiation incidence, is considered. The directions of incidence for the parallel beams are anterior-posterior, posterior-anterior, left lateral, right lateral and a full 360 rotation around the body`s longitudinal axis. The influence of beam divergence on the body doses is also considered as well as the dependence of effective dose on the angle of radiation incidence. Regarding exposure of the public to environmental sources, three source geometries are considered: exposure from a radioactive cloud, from ground contamination and from the natural radionuclides distributed homogeneously in the ground. The precise angular and energy distributions of the gamma rays incident on the human body were taken into account. The organ dose conversion coefficients given in this catalogue were calculated using a Monte Carlo code simulating the photon transport in mathematical models of an adult male and an adult female, respectively. Conversion coefficients are given for the equivalent dose of 23 organs and tissues as well as for effective dose and the equivalent dose of the so-called `remainder`. The organ equivalent dose conversion coefficients are given separately for the adult male and female models and - as arithmetic mean of the conversion coefficients of both - for an average adult. Fitted data of the coefficients are presented in tables; the primary raw data as resulting from the Monte Carlo calculation are shown in figures together with the fitted data. (orig.)

The calculation of dose from external photon exposures using reference human phantoms and Monte Carlo methods. Pt. 7. Organ doses due to parallel and environmental exposure geometries

International Nuclear Information System (INIS)

Zankl, M.

1997-03-01

This report presents a tabulation of organ and tissue equivalent dose as well as effective dose conversion coefficients, normalised to air kerma free in air, for occupational exposures and environmental exposures of the public to external photon radiation. For occupational exposures, whole-body irradiation with idealised geometries, i.e. broad parallel beams and fully isotropic radiation incidence, is considered. The directions of incidence for the parallel beams are anterior-posterior, posterior-anterior, left lateral, right lateral and a full 360 rotation around the body's longitudinal axis. The influence of beam divergence on the body doses is also considered as well as the dependence of effective dose on the angle of radiation incidence. Regarding exposure of the public to environmental sources, three source geometries are considered: exposure from a radioactive cloud, from ground contamination and from the natural radionuclides distributed homogeneously in the ground. The precise angular and energy distributions of the gamma rays incident on the human body were taken into account. The organ dose conversion coefficients given in this catalogue were calculated using a Monte Carlo code simulating the photon transport in mathematical models of an adult male and an adult female, respectively. Conversion coefficients are given for the equivalent dose of 23 organs and tissues as well as for effective dose and the equivalent dose of the so-called 'remainder'. The organ equivalent dose conversion coefficients are given separately for the adult male and female models and - as arithmetic mean of the conversion coefficients of both - for an average adult. Fitted data of the coefficients are presented in tables; the primary raw data as resulting from the Monte Carlo calculation are shown in figures together with the fitted data. (orig.)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Monte Carlo simulation of a varian 21EX Clinac 6 MV photon beam characteristics using GATE6

Energy Technology Data Exchange (ETDEWEB)

An, Jung Su [Korea Atomic Energy Research Institute, Nonproliferation System Research Division, Daejeon (Korea, Republic of); Lee, Chang Lae [Center for Radiological Environment and Health Science, Busan (Korea, Republic of); Baek, Cheol Ha [Dept. of Radiological Science, College of Health Science, Yonsei University, Seoul (Korea, Republic of)

2016-12-15

Monte Carlo simulations are widely used as the most accurate technique for dose calculation in radiation therapy. In this paper, the GATE6(Geant4 Application for Tomographic Emission ver.6) code was employed to calculate the dosimetric performance of the photon beams from a linear accelerator(LINAC). The treatment head of a Varian 21EX Clinac was modeled including the major geometric structures within the beam path such as a target, a primary collimator, a flattening filter, a ion chamber, and jaws. The 6 MV photon spectra were characterized in a standard 10 x 10 cm{sup 2} field at 100 cm source-to-surface distance(SSD) and subsequent dose estimations were made in a water phantom. The measurements of percentage depth dose and dose profiles were performed with 3D water phantom and the simulated data was compared to measured reference data. The simulated results agreed very well with the measured data. It has been found that the GATE6 code is an effective tool for dose optimization in radiotherapy applications.

Monte Carlo and analytic simulations in nanoparticle-enhanced radiation therapy

Directory of Open Access Journals (Sweden)

Paro AD

2016-09-01

Full Text Available Autumn D Paro,1 Mainul Hossain,2 Thomas J Webster,1,3,4 Ming Su1,4 1Department of Chemical Engineering, Northeastern University, Boston, MA, USA; 2NanoScience Technology Center and School of Electrical Engineering and Computer Science, University of Central Florida, Orlando, Florida, USA; 3Excellence for Advanced Materials Research, King Abdulaziz University, Jeddah, Saudi Arabia; 4Wenzhou Institute of Biomaterials and Engineering, Chinese Academy of Science, Wenzhou Medical University, Zhejiang, People’s Republic of China Abstract: Analytical and Monte Carlo simulations have been used to predict dose enhancement factors in nanoparticle-enhanced X-ray radiation therapy. Both simulations predict an increase in dose enhancement in the presence of nanoparticles, but the two methods predict different levels of enhancement over the studied energy, nanoparticle materials, and concentration regime for several reasons. The Monte Carlo simulation calculates energy deposited by electrons and photons, while the analytical one only calculates energy deposited by source photons and photoelectrons; the Monte Carlo simulation accounts for electron–hole recombination, while the analytical one does not; and the Monte Carlo simulation randomly samples photon or electron path and accounts for particle interactions, while the analytical simulation assumes a linear trajectory. This study demonstrates that the Monte Carlo simulation will be a better choice to evaluate dose enhancement with nanoparticles in radiation therapy. Keywords: nanoparticle, dose enhancement, Monte Carlo simulation, analytical simulation, radiation therapy, tumor cell, X-ray 

Applications of the Los Alamos High Energy Transport code

International Nuclear Information System (INIS)

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

1992-01-01

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

Monte Carlo methods for flux expansion solutions of transport problems

International Nuclear Information System (INIS)

Spanier, J.

1999-01-01

Adaptive Monte Carlo methods, based on the use of either correlated sampling or importance sampling, to obtain global solutions to certain transport problems have recently been described. The resulting learning algorithms are capable of achieving geometric convergence when applied to the estimation of a finite number of coefficients in a flux expansion representation of the global solution. However, because of the nonphysical nature of the random walk simulations needed to perform importance sampling, conventional transport estimators and source sampling techniques require modification to be used successfully in conjunction with such flux expansion methods. It is shown how these problems can be overcome. First, the traditional path length estimators in wide use in particle transport simulations are generalized to include rather general detector functions (which, in this application, are the individual basis functions chosen for the flus expansion). Second, it is shown how to sample from the signed probabilities that arise as source density functions in these applications, without destroying the zero variance property needed to ensure geometric convergence to zero error

The electron transport problem sampling by Monte Carlo individual collision technique

Energy Technology Data Exchange (ETDEWEB)

Androsenko, P.A.; Belousov, V.I. [Obninsk State Technical Univ. of Nuclear Power Engineering, Kaluga region (Russian Federation)

2005-07-01

The problem of electron transport is of most interest in all fields of the modern science. To solve this problem the Monte Carlo sampling has to be used. The electron transport is characterized by a large number of individual interactions. To simulate electron transport the 'condensed history' technique may be used where a large number of collisions are grouped into a single step to be sampled randomly. Another kind of Monte Carlo sampling is the individual collision technique. In comparison with condensed history technique researcher has the incontestable advantages. For example one does not need to give parameters altered by condensed history technique like upper limit for electron energy, resolution, number of sub-steps etc. Also the condensed history technique may lose some very important tracks of electrons because of its limited nature by step parameters of particle movement and due to weakness of algorithms for example energy indexing algorithm. There are no these disadvantages in the individual collision technique. This report presents some sampling algorithms of new version BRAND code where above mentioned technique is used. All information on electrons was taken from Endf-6 files. They are the important part of BRAND. These files have not been processed but directly taken from electron information source. Four kinds of interaction like the elastic interaction, the Bremsstrahlung, the atomic excitation and the atomic electro-ionization were considered. In this report some results of sampling are presented after comparison with analogs. For example the endovascular radiotherapy problem (P2) of QUADOS2002 was presented in comparison with another techniques that are usually used. (authors)

Guiding electromagnetic waves around sharp corners: topologically protected photonic transport in meta-waveguides (Presentation Recording)

Science.gov (United States)

Shvets, Gennady B.; Khanikaev, Alexander B.; Ma, Tzuhsuan; Lai, Kueifu

2015-09-01

Science thrives on analogies, and a considerable number of inventions and discoveries have been made by pursuing an unexpected connection to a very different field of inquiry. For example, photonic crystals have been referred to as "semiconductors of light" because of the far-reaching analogies between electron propagation in a crystal lattice and light propagation in a periodically modulated photonic environment. However, two aspects of electron behavior, its spin and helicity, escaped emulation by photonic systems until recent invention of photonic topological insulators (PTIs). The impetus for these developments in photonics came from the discovery of topologically nontrivial phases in condensed matter physics enabling edge states immune to scattering. The realization of topologically protected transport in photonics would circumvent a fundamental limitation imposed by the wave equation: inability of reflections-free light propagation along sharply bent pathway. Topologically protected electromagnetic states could be used for transporting photons without any scattering, potentially underpinning new revolutionary concepts in applied science and engineering. I will demonstrate that a PTI can be constructed by applying three types of perturbations: (a) finite bianisotropy, (b) gyromagnetic inclusion breaking the time-reversal (T) symmetry, and (c) asymmetric rods breaking the parity (P) symmetry. We will experimentally demonstrate (i) the existence of the full topological bandgap in a bianisotropic, and (ii) the reflectionless nature of wave propagation along the interface between two PTIs with opposite signs of the bianisotropy.

Recent developments of JAEA's Monte Carlo Code MVP for reactor physics applications

International Nuclear Information System (INIS)

Nagaya, Y.; Okumura, K.; Mori, T.

2013-01-01

MVP is a general-purpose continuous-energy Monte Carlo code for neutron and photon transport calculations that has been developed since the late 1980's at Japan Atomic Energy Agency (JAEA, formerly JAERI). The MVP code is designed for nuclear reactor applications such as reactor core design/analysis, criticality safety and reactor shielding. This paper describes the MVP code and present its latest developments. Among the new capabilities of MVP we find: -) the perturbation method has been implemented for the change in k(eff); -) the eigenvalue calculations can be performed with an explicit treatment of delayed neutrons in which their fission spectra are taken into account; -) the capability of tallying the scattering matrix (group-to-group scattering cross sections); -) the implementation of an exact model for resonance elastic scattering; and -) a Monte Carlo perturbation technique is used to calculate reactor kinetics parameters

Parallel Monte Carlo Particle Transport and the Quality of Random Number Generators: How Good is Good Enough?

International Nuclear Information System (INIS)

Procassini, R J; Beck, B R

2004-01-01

It might be assumed that use of a ''high-quality'' random number generator (RNG), producing a sequence of ''pseudo random'' numbers with a ''long'' repetition period, is crucial for producing unbiased results in Monte Carlo particle transport simulations. While several theoretical and empirical tests have been devised to check the quality (randomness and period) of an RNG, for many applications it is not clear what level of RNG quality is required to produce unbiased results. This paper explores the issue of RNG quality in the context of parallel, Monte Carlo transport simulations in order to determine how ''good'' is ''good enough''. This study employs the MERCURY Monte Carlo code, which incorporates the CNPRNG library for the generation of pseudo-random numbers via linear congruential generator (LCG) algorithms. The paper outlines the usage of random numbers during parallel MERCURY simulations, and then describes the source and criticality transport simulations which comprise the empirical basis of this study. A series of calculations for each test problem in which the quality of the RNG (period of the LCG) is varied provides the empirical basis for determining the minimum repetition period which may be employed without producing a bias in the mean integrated results

Determination of output factor for 6 MV small photon beam: comparison between Monte Carlo simulation technique and microDiamond detector

International Nuclear Information System (INIS)

Krongkietlearts, K; Tangboonduangjit, P; Paisangittisakul, N

2016-01-01

In order to improve the life's quality for a cancer patient, the radiation techniques are constantly evolving. Especially, the two modern techniques which are intensity modulated radiation therapy (IMRT) and volumetric modulated arc therapy (VMAT) are quite promising. They comprise of many small beam sizes (beamlets) with various intensities to achieve the intended radiation dose to the tumor and minimal dose to the nearby normal tissue. The study investigates whether the microDiamond detector (PTW manufacturer), a synthetic single crystal diamond detector, is suitable for small field output factor measurement. The results were compared with those measured by the stereotactic field detector (SFD) and the Monte Carlo simulation (EGSnrc/BEAMnrc/DOSXYZ). The calibration of Monte Carlo simulation was done using the percentage depth dose and dose profile measured by the photon field detector (PFD) of the 10×10 cm 2 field size with 100 cm SSD. Comparison of the values obtained from the calculations and measurements are consistent, no more than 1% difference. The output factors obtained from the microDiamond detector have been compared with those of SFD and Monte Carlo simulation, the results demonstrate the percentage difference of less than 2%. (paper)

Optimal Spatial Subdivision method for improving geometry navigation performance in Monte Carlo particle transport simulation

International Nuclear Information System (INIS)

Chen, Zhenping; Song, Jing; Zheng, Huaqing; Wu, Bin; Hu, Liqin

2015-01-01

Highlights: • The subdivision combines both advantages of uniform and non-uniform schemes. • The grid models were proved to be more efficient than traditional CSG models. • Monte Carlo simulation performance was enhanced by Optimal Spatial Subdivision. • Efficiency gains were obtained for realistic whole reactor core models. - Abstract: Geometry navigation is one of the key aspects of dominating Monte Carlo particle transport simulation performance for large-scale whole reactor models. In such cases, spatial subdivision is an easily-established and high-potential method to improve the run-time performance. In this study, a dedicated method, named Optimal Spatial Subdivision, is proposed for generating numerically optimal spatial grid models, which are demonstrated to be more efficient for geometry navigation than traditional Constructive Solid Geometry (CSG) models. The method uses a recursive subdivision algorithm to subdivide a CSG model into non-overlapping grids, which are labeled as totally or partially occupied, or not occupied at all, by CSG objects. The most important point is that, at each stage of subdivision, a conception of quality factor based on a cost estimation function is derived to evaluate the qualities of the subdivision schemes. Only the scheme with optimal quality factor will be chosen as the final subdivision strategy for generating the grid model. Eventually, the model built with the optimal quality factor will be efficient for Monte Carlo particle transport simulation. The method has been implemented and integrated into the Super Monte Carlo program SuperMC developed by FDS Team. Testing cases were used to highlight the performance gains that could be achieved. Results showed that Monte Carlo simulation runtime could be reduced significantly when using the new method, even as cases reached whole reactor core model sizes

Comparison of preconditioned generalized conjugate gradient methods to two-dimensional neutron and photon transport equation

International Nuclear Information System (INIS)

Chen, G.S.

1997-01-01

We apply and compare the preconditioned generalized conjugate gradient methods to solve the linear system equation that arises in the two-dimensional neutron and photon transport equation in this paper. Several subroutines are developed on the basis of preconditioned generalized conjugate gradient methods for time-independent, two-dimensional neutron and photon transport equation in the transport theory. These generalized conjugate gradient methods are used. TFQMR (transpose free quasi-minimal residual algorithm), CGS (conjuage gradient square algorithm), Bi-CGSTAB (bi-conjugate gradient stabilized algorithm) and QMRCGSTAB (quasi-minimal residual variant of bi-conjugate gradient stabilized algorithm). These sub-routines are connected to computer program DORT. Several problems are tested on a personal computer with Intel Pentium CPU. (author)

Dirac tensor with heavy photon

Energy Technology Data Exchange (ETDEWEB)

Bytev, V.V.; Kuraev, E.A. [Joint Institute of Nuclear Research, Moscow (Russian Federation). Bogoliubov Lab. of Theoretical Physics; Scherbakova, E.S. [Hamburg Univ. (Germany). 1. Inst. fuer Theoretische Physik

2012-01-15

For the large-angles hard photon emission by initial leptons in process of high energy annihilation of e{sup +}e{sup -} {yields} to hadrons the Dirac tensor is obtained, taking into account the lowest order radiative corrections. The case of large-angles emission of two hard photons by initial leptons is considered. This result is being completed by the kinematics case of collinear hard photons emission as well as soft virtual and real photons and can be used for construction of Monte-Carlo generators. (orig.)

CAD-based Monte Carlo program for integrated simulation of nuclear system SuperMC

International Nuclear Information System (INIS)

Wu, Y.; Song, J.; Zheng, H.; Sun, G.; Hao, L.; Long, P.; Hu, L.

2013-01-01

SuperMC is a (Computer-Aided-Design) CAD-based Monte Carlo (MC) program for integrated simulation of nuclear systems developed by FDS Team (China), making use of hybrid MC-deterministic method and advanced computer technologies. The design aim, architecture and main methodology of SuperMC are presented in this paper. The taking into account of multi-physics processes and the use of advanced computer technologies such as automatic geometry modeling, intelligent data analysis and visualization, high performance parallel computing and cloud computing, contribute to the efficiency of the code. SuperMC2.1, the latest version of the code for neutron, photon and coupled neutron and photon transport calculation, has been developed and validated by using a series of benchmarking cases such as the fusion reactor ITER model and the fast reactor BN-600 model

Error reduction techniques for Monte Carlo neutron transport calculations

International Nuclear Information System (INIS)

Ju, J.H.W.

1981-01-01