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Sample records for carlo transport code

  1. THE MCNPX MONTE CARLO RADIATION TRANSPORT CODE

    Energy Technology Data Exchange (ETDEWEB)

    WATERS, LAURIE S. [Los Alamos National Laboratory; MCKINNEY, GREGG W. [Los Alamos National Laboratory; DURKEE, JOE W. [Los Alamos National Laboratory; FENSIN, MICHAEL L. [Los Alamos National Laboratory; JAMES, MICHAEL R. [Los Alamos National Laboratory; JOHNS, RUSSELL C. [Los Alamos National Laboratory; PELOWITZ, DENISE B. [Los Alamos National Laboratory

    2007-01-10

    MCNPX (Monte Carlo N-Particle eXtended) is a general-purpose Monte Carlo radiation transport code with three-dimensional geometry and continuous-energy transport of 34 particles and light ions. It contains flexible source and tally options, interactive graphics, and support for both sequential and multi-processing computer platforms. MCNPX is based on MCNP4B, and has been upgraded to most MCNP5 capabilities. MCNP is a highly stable code tracking neutrons, photons and electrons, and using evaluated nuclear data libraries for low-energy interaction probabilities. MCNPX has extended this base to a comprehensive set of particles and light ions, with heavy ion transport in development. Models have been included to calculate interaction probabilities when libraries are not available. Recent additions focus on the time evolution of residual nuclei decay, allowing calculation of transmutation and delayed particle emission. MCNPX is now a code of great dynamic range, and the excellent neutronics capabilities allow new opportunities to simulate devices of interest to experimental particle physics; particularly calorimetry. This paper describes the capabilities of the current MCNPX version 2.6.C, and also discusses ongoing code development.

  2. Morse Monte Carlo Radiation Transport Code System

    Energy Technology Data Exchange (ETDEWEB)

    Emmett, M.B.

    1975-02-01

    The report contains sections containing descriptions of the MORSE and PICTURE codes, input descriptions, sample problems, deviations of the physical equations and explanations of the various error messages. The MORSE code is a multipurpose neutron and gamma-ray transport Monte Carlo code. Time dependence for both shielding and criticality problems is provided. General three-dimensional geometry may be used with an albedo option available at any material surface. The PICTURE code provide aid in preparing correct input data for the combinatorial geometry package CG. It provides a printed view of arbitrary two-dimensional slices through the geometry. By inspecting these pictures one may determine if the geometry specified by the input cards is indeed the desired geometry. 23 refs. (WRF)

  3. 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.)

  4. TRIPOLI-3: a neutron/photon Monte Carlo transport code

    Energy Technology Data Exchange (ETDEWEB)

    Nimal, J.C.; Vergnaud, T. [Commissariat a l' Energie Atomique, Gif-sur-Yvette (France). Service d' Etudes de Reacteurs et de Mathematiques Appliquees

    2001-07-01

    The present version of TRIPOLI-3 solves the transport equation for coupled neutron and gamma ray problems in three dimensional geometries by using the Monte Carlo method. This code is devoted both to shielding and criticality problems. The most important feature for particle transport equation solving is the fine treatment of the physical phenomena and sophisticated biasing technics useful for deep penetrations. The code is used either for shielding design studies or for reference and benchmark to validate cross sections. Neutronic studies are essentially cell or small core calculations and criticality problems. TRIPOLI-3 has been used as reference method, for example, for resonance self shielding qualification. (orig.)

  5. Parallelization of a Monte Carlo particle transport simulation code

    Science.gov (United States)

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

    2010-05-01

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

  6. 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...

  7. MORSE Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Cramer, S.N.

    1984-01-01

    The MORSE code is a large general-use multigroup Monte Carlo code system. Although no claims can be made regarding its superiority in either theoretical details or Monte Carlo techniques, MORSE has been, since its inception at ORNL in the late 1960s, the most widely used Monte Carlo radiation transport code. The principal reason for this popularity is that MORSE is relatively easy to use, independent of any installation or distribution center, and it can be easily customized to fit almost any specific need. Features of the MORSE code are described.

  8. Neutral Particle Transport in Cylindrical Plasma Simulated by a Monte Carlo Code

    Institute of Scientific and Technical Information of China (English)

    YU Deliang; YAN Longwen; ZHONG Guangwu; LU Jie; YI Ping

    2007-01-01

    A Monte Carlo code (MCHGAS) has been developed to investigate the neutral particle transport.The code can calculate the radial profile and energy spectrum of neutral particles in cylindrical plasmas.The calculation time of the code is dramatically reduced when the Splitting and Roulette schemes are applied. The plasma model of an infinite cylinder is assumed in the code,which is very convenient in simulating neutral particle transports in small and middle-sized tokamaks.The design of the multi-channel neutral particle analyser (NPA) on HL-2A can be optimized by using this code.

  9. TART97 a coupled neutron-photon 3-D, combinatorial geometry Monte Carlo transport code

    Energy Technology Data Exchange (ETDEWEB)

    Cullen, D.E.

    1997-11-22

    TART97 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo transport code. This code can on any modern computer. It is a complete system to assist you with input preparation, running Monte Carlo calculations, and analysis of output results. TART97 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. TART97 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. TART97 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART97 and its data riles.

  10. 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...

  11. Update on the Development and Validation of MERCURY: A Modern, Monte Carlo Particle Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Procassini, R J; Taylor, J M; McKinley, M S; Greenman, G M; Cullen, D E; O' Brien, M J; Beck, B R; Hagmann, C A

    2005-06-06

    An update on the development and validation of the MERCURY Monte Carlo particle transport code is presented. MERCURY is a modern, parallel, general-purpose Monte Carlo code being developed at the Lawrence Livermore National Laboratory. During the past year, several major algorithm enhancements have been completed. These include the addition of particle trackers for 3-D combinatorial geometry (CG), 1-D radial meshes, 2-D quadrilateral unstructured meshes, as well as a feature known as templates for defining recursive, repeated structures in CG. New physics capabilities include an elastic-scattering neutron thermalization model, support for continuous energy cross sections and S ({alpha}, {beta}) molecular bound scattering. Each of these new physics features has been validated through code-to-code comparisons with another Monte Carlo transport code. Several important computer science features have been developed, including an extensible input-parameter parser based upon the XML data description language, and a dynamic load-balance methodology for efficient parallel calculations. This paper discusses the recent work in each of these areas, and describes a plan for future extensions that are required to meet the needs of our ever expanding user base.

  12. A fast Monte Carlo code for proton transport in radiation therapy based on MCNPX.

    Science.gov (United States)

    Jabbari, Keyvan; Seuntjens, Jan

    2014-07-01

    An important requirement for proton therapy is a software for dose calculation. Monte Carlo is the most accurate method for dose calculation, but it is very slow. In this work, a method is developed to improve the speed of dose calculation. The method is based on pre-generated tracks for particle transport. The MCNPX code has been used for generation of tracks. A set of data including the track of the particle was produced in each particular material (water, air, lung tissue, bone, and soft tissue). This code can transport protons in wide range of energies (up to 200 MeV for proton). The validity of the fast Monte Carlo (MC) code is evaluated with data MCNPX as a reference code. While analytical pencil beam algorithm transport shows great errors (up to 10%) near small high density heterogeneities, there was less than 2% deviation of MCNPX results in our dose calculation and isodose distribution. In terms of speed, the code runs 200 times faster than MCNPX. In the Fast MC code which is developed in this work, it takes the system less than 2 minutes to calculate dose for 10(6) particles in an Intel Core 2 Duo 2.66 GHZ desktop computer.

  13. A fast Monte Carlo code for proton transport in radiation therapy based on MCNPX

    Directory of Open Access Journals (Sweden)

    Keyvan Jabbari

    2014-01-01

    Full Text Available An important requirement for proton therapy is a software for dose calculation. Monte Carlo is the most accurate method for dose calculation, but it is very slow. In this work, a method is developed to improve the speed of dose calculation. The method is based on pre-generated tracks for particle transport. The MCNPX code has been used for generation of tracks. A set of data including the track of the particle was produced in each particular material (water, air, lung tissue, bone, and soft tissue. This code can transport protons in wide range of energies (up to 200 MeV for proton. The validity of the fast Monte Carlo (MC code is evaluated with data MCNPX as a reference code. While analytical pencil beam algorithm transport shows great errors (up to 10% near small high density heterogeneities, there was less than 2% deviation of MCNPX results in our dose calculation and isodose distribution. In terms of speed, the code runs 200 times faster than MCNPX. In the Fast MC code which is developed in this work, it takes the system less than 2 minutes to calculate dose for 10 6 particles in an Intel Core 2 Duo 2.66 GHZ desktop computer.

  14. 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.

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

    CERN Document Server

    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...

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

    Science.gov (United States)

    Rabie, M.; Franck, C. M.

    2016-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.R.; Cummings, J.C. [Los Alamos National Lab., NM (United States); Nolen, S.D. [Texas A and M Univ., College Station, TX (United States)]|[Los Alamos National Lab., NM (United States)

    1997-05-01

    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 {alpha}-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 {alpha}-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed.

  18. Verification of Three Dimensional Triangular Prismatic Discrete Ordinates Transport Code ENSEMBLE-TRIZ by Comparison with Monte Carlo Code GMVP

    Science.gov (United States)

    Homma, Yuto; Moriwaki, Hiroyuki; Ohki, Shigeo; Ikeda, Kazumi

    2014-06-01

    This paper deals with verification of three dimensional triangular prismatic discrete ordinates transport calculation code ENSEMBLE-TRIZ by comparison with multi-group Monte Carlo calculation code GMVP in a large fast breeder reactor. The reactor is a 750 MWe electric power sodium cooled reactor. Nuclear characteristics are calculated at beginning of cycle of an initial core and at beginning and end of cycle of equilibrium core. According to the calculations, the differences between the two methodologies are smaller than 0.0002 Δk in the multi-plication factor, relatively about 1% in the control rod reactivity, and 1% in the sodium void reactivity.

  19. Neutron cross-section probability tables in TRIPOLI-3 Monte Carlo transport code

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, S.H.; Vergnaud, T.; Nimal, J.C. [Commissariat a l`Energie Atomique, Gif-sur-Yvette (France). Lab. d`Etudes de Protection et de Probabilite

    1998-03-01

    Neutron transport calculations need an accurate treatment of cross sections. Two methods (multi-group and pointwise) are usually used. A third one, the probability table (PT) method, has been developed to produce a set of cross-section libraries, well adapted to describe the neutron interaction in the unresolved resonance energy range. Its advantage is to present properly the neutron cross-section fluctuation within a given energy group, allowing correct calculation of the self-shielding effect. Also, this PT cross-section representation is suitable for simulation of neutron propagation by the Monte Carlo method. The implementation of PTs in the TRIPOLI-3 three-dimensional general Monte Carlo transport code, developed at Commissariat a l`Energie Atomique, and several validation calculations are presented. The PT method is proved to be valid not only in the unresolved resonance range but also in all the other energy ranges.

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

    CERN Document Server

    Ilic, R D; Stankovic, S J

    2002-01-01

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

  1. 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.

  2. A fast Monte Carlo code for proton transport in radiation therapy based on MCNPX

    OpenAIRE

    Keyvan Jabbari; Jan Seuntjens

    2014-01-01

    An important requirement for proton therapy is a software for dose calculation. Monte Carlo is the most accurate method for dose calculation, but it is very slow. In this work, a method is developed to improve the speed of dose calculation. The method is based on pre-generated tracks for particle transport. The MCNPX code has been used for generation of tracks. A set of data including the track of the particle was produced in each particular material (water, air, lung tissue, bone, and soft t...

  3. Domain Decomposition of a Constructive Solid Geometry Monte Carlo Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    O' Brien, M J; Joy, K I; Procassini, R J; Greenman, G M

    2008-12-07

    Domain decomposition has been implemented in a Constructive Solid Geometry (CSG) Monte Carlo neutron transport code. Previous methods to parallelize a CSG code relied entirely on particle parallelism; but in our approach we distribute the geometry as well as the particles across processors. This enables calculations whose geometric description is larger than what could fit in memory of a single processor, thus it must be distributed across processors. In addition to enabling very large calculations, we show that domain decomposition can speed up calculations compared to particle parallelism alone. We also show results of a calculation of the proposed Laser Inertial-Confinement Fusion-Fission Energy (LIFE) facility, which has 5.6 million CSG parts.

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

    Energy Technology Data Exchange (ETDEWEB)

    Higuchi, Kenji; Takemiya, Hiroshi [Japan Atomic Energy Research Inst., Tokyo (Japan); Kawasaki, Takuji [Fuji Research Institute Corporation, Tokyo (Japan)

    2001-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Kensek, R.P.; Lorence, L.J.; Halbleib, J.A. [Sandia National Labs., Albuquerque, NM (United States); Morel, J.E. [Los Alamos National Lab., NM (United States)

    1996-07-01

    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.

  6. TART98 a coupled neutron-photon 3-D, combinatorial geometry time dependent Monte Carlo Transport code

    Energy Technology Data Exchange (ETDEWEB)

    Cullen, D E

    1998-11-22

    TART98 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. TART98 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. TART98 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. TART98 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART98 and its data files.

  7. 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)

  8. TRIPOLI: a general Monte Carlo code, present state and future prospects. [Neutron and gamma ray transport

    Energy Technology Data Exchange (ETDEWEB)

    Nimal, J.C.; Vergnaud, T. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France))

    1990-01-01

    This paper describes the most important features of the Monte Carlo code TRIPOLI-2. This code solves the Boltzmann equation in three-dimensional geometries for coupled neutron and gamma rays problems. A particular emphasis is devoted to the biasing techniques, which are very important for deep penetration. Future developments in TRIPOLI are described in the conclusion. (author).

  9. Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

    2015-07-01

    Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

  10. The application of the Monte-Carlo neutron transport code MCNP to a small "nuclear battery" system

    OpenAIRE

    Puigdellívol Sadurní, Roger

    2009-01-01

    The project consist in calculate the keff to a small nuclear battery. The code Monte- Carlo neutron transport code MCNP is used to calculate the keff. The calculations are done at the beginning of life to know the capacity of the core becomes critical in different conditions. These conditions are the study parameters that determine the criticality of the core. These parameters are the uranium enrichment, the coated particles (TRISO) packing factor and the size of the core. More...

  11. ITS version 5.0 : 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

    2004-06-01

    ITS is a powerful and user-friendly software package permitting state of the art Monte Carlo solution of linear time-independent couple 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, and (3) parallel implementations of all ITS codes. Moreover the general user friendliness of the software has been enhanced through increased internal error checking and improved code portability.

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

    Science.gov (United States)

    Schnittman, Jeremy David; Krolik, Julian H.

    2013-01-01

    We present a new code for radiation transport around Kerr black holes, including arbitrary emission and absorption mechanisms, as well as electron scattering and polarization. The code is particularly useful for analyzing accretion flows made up of optically thick disks and optically thin coronae. We give a detailed description of the methods employed in the code and also present results from a number of numerical tests to assess its accuracy and convergence.

  13. Update on the Status of the FLUKA Monte Carlo Transport Code

    Science.gov (United States)

    Pinsky, L.; Anderson, V.; Empl, A.; Lee, K.; Smirnov, G.; Zapp, N; Ferrari, A.; Tsoulou, K.; Roesler, S.; Vlachoudis, V.; Battisoni, G.; Ceruti, F.; Gadioli, M. V.; Garzelli, M.; Muraro, S.; Rancati, T.; Sala, P.; Ballarini, R.; Ottolenghi, A.; Parini, V.; Scannicchio, D.; Pelliccioni, M.; Wilson, T. L.

    2004-01-01

    The FLUKA Monte Carlo transport code is a well-known simulation tool in High Energy Physics. FLUKA is a dynamic tool in the sense that it is being continually updated and improved by the authors. Here we review the progresses achieved in the last year on the physics models. From the point of view of hadronic physics, most of the effort is still in the field of nucleus--nucleus interactions. The currently available version of FLUKA already includes the internal capability to simulate inelastic nuclear interactions beginning with lab kinetic energies of 100 MeV/A up the the highest accessible energies by means of the DPMJET-II.5 event generator to handle the interactions for greater than 5 GeV/A and rQMD for energies below that. The new developments concern, at high energy, the embedding of the DPMJET-III generator, which represent a major change with respect to the DPMJET-II structure. This will also allow to achieve a better consistency between the nucleus-nucleus section with the original FLUKA model for hadron-nucleus collisions. Work is also in progress to implement a third event generator model based on the Master Boltzmann Equation approach, in order to extend the energy capability from 100 MeV/A down to the threshold for these reactions. In addition to these extended physics capabilities, structural changes to the programs input and scoring capabilities are continually being upgraded. In particular we want to mention the upgrades in the geometry packages, now capable of reaching higher levels of abstraction. Work is also proceeding to provide direct import into ROOT of the FLUKA output files for analysis and to deploy a user-friendly GUI input interface.

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

    Energy Technology Data Exchange (ETDEWEB)

    Taasti, Vicki Trier; Knudsen, Helge [Dept. of Physics and Astronomy, Aarhus University (Denmark); Holzscheiter, Michael H. [Dept. of Physics and Astronomy, Aarhus University (Denmark); Dept. of Physics and Astronomy, University of New Mexico (United States); Sobolevsky, Nikolai [Institute for Nuclear Research of the Russian Academy of Sciences (INR), Moscow (Russian Federation); Moscow Institute of Physics and Technology (MIPT), Dolgoprudny (Russian Federation); Thomsen, Bjarne [Dept. of Physics and Astronomy, Aarhus University (Denmark); Bassler, Niels, E-mail: bassler@phys.au.dk [Dept. of Physics and Astronomy, Aarhus University (Denmark)

    2015-03-15

    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 experimental depth dose curve obtained by the AD-4/ACE collaboration was compared with an earlier version of SHIELD-HIT, but since then inelastic annihilation cross sections for antiprotons have been updated and a more detailed geometric model of the AD-4/ACE experiment was applied. Furthermore, the Fermi–Teller Z-law, which is implemented by default in SHIELD-HIT12A has been shown not to be a good approximation for the capture probability of negative projectiles by nuclei. We investigate other theories which have been developed, and give a better agreement with experimental findings. The consequence of these updates is tested by comparing simulated data with the antiproton depth dose curve in water. It is found that the implementation of these new capture probabilities results in an overestimation of the depth dose curve in the Bragg peak. This can be mitigated by scaling the antiproton collision cross sections, which restores the agreement, but some small deviations still remain. Best agreement is achieved by using the most recent antiproton collision cross sections and the Fermi–Teller Z-law, even if experimental data conclude that the Z-law is inadequately describing annihilation on compounds. We conclude that more experimental cross section data are needed in the lower energy range in order to resolve this contradiction, ideally combined with more rigorous models for annihilation on compounds.

  15. TRIPOLI capabilities proved by a set of solved problems. [Monte Carlo neutron and gamma ray transport code

    Energy Technology Data Exchange (ETDEWEB)

    Vergnaud, T.; Nimal, J.C. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France))

    1990-01-01

    The three-dimensional polycinetic Monte Carlo particle transport code TRIPOLI has been under development in the French Shielding Laboratory at Saclay since 1965. TRIPOLI-1 began to run in 1970 and became TRIPOLI-2 in 1978: since then its capabilities have been improved and many studies have been performed. TRIPOLI can treat stationary or time dependent problems in shielding and in neutronics. Some examples of solved problems are presented to demonstrate the many possibilities of the system. (author).

  16. Generation of discrete scattering cross sections and demonstration of Monte Carlo charged particle transport in the Milagro IMC code package

    Energy Technology Data Exchange (ETDEWEB)

    Walsh, J. A. [Department of Nuclear Science and Engineering, Massachusetts Institute of Technology, NW12-312 Albany, St. Cambridge, MA 02139 (United States); Palmer, T. S. [Department of Nuclear Engineering and Radiation Health Physics, Oregon State University, 116 Radiation Center, Corvallis, OR 97331 (United States); Urbatsch, T. J. [XTD-5: Air Force Systems, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2013-07-01

    A new method for generating discrete scattering cross sections to be used in charged particle transport calculations is investigated. The method of data generation is presented and compared to current methods for obtaining discrete cross sections. The new, more generalized approach allows greater flexibility in choosing a cross section model from which to derive discrete values. Cross section data generated with the new method is verified through a comparison with discrete data obtained with an existing method. Additionally, a charged particle transport capability is demonstrated in the time-dependent Implicit Monte Carlo radiative transfer code package, Milagro. The implementation of this capability is verified using test problems with analytic solutions as well as a comparison of electron dose-depth profiles calculated with Milagro and an already-established electron transport code. An initial investigation of a preliminary integration of the discrete cross section generation method with the new charged particle transport capability in Milagro is also presented. (authors)

  17. Parallel Monte Carlo transport modeling in the context of a time-dependent, three-dimensional multi-physics code

    Energy Technology Data Exchange (ETDEWEB)

    Procassini, R.J. [Lawrence Livermore National lab., CA (United States)

    1997-12-31

    The fine-scale, multi-space resolution that is envisioned for accurate simulations of complex weapons systems in three spatial dimensions implies flop-rate and memory-storage requirements that will only be obtained in the near future through the use of parallel computational techniques. Since the Monte Carlo transport models in these simulations usually stress both of these computational resources, they are prime candidates for parallelization. The MONACO Monte Carlo transport package, which is currently under development at LLNL, will utilize two types of parallelism within the context of a multi-physics design code: decomposition of the spatial domain across processors (spatial parallelism) and distribution of particles in a given spatial subdomain across additional processors (particle parallelism). This implementation of the package will utilize explicit data communication between domains (message passing). Such a parallel implementation of a Monte Carlo transport model will result in non-deterministic communication patterns. The communication of particles between subdomains during a Monte Carlo time step may require a significant level of effort to achieve a high parallel efficiency.

  18. Towards scalable parellelism in Monte Carlo particle transport codes using remote memory access

    Energy Technology Data Exchange (ETDEWEB)

    Romano, Paul K [Los Alamos National Laboratory; Brown, Forrest B [Los Alamos National Laboratory; Forget, Benoit [MIT

    2010-01-01

    One forthcoming challenge in the area of high-performance computing is having the ability to run large-scale problems while coping with less memory per compute node. In this work, they investigate a novel data decomposition method that would allow Monte Carlo transport calculations to be performed on systems with limited memory per compute node. In this method, each compute node remotely retrieves a small set of geometry and cross-section data as needed and remotely accumulates local tallies when crossing the boundary of the local spatial domain. initial results demonstrate that while the method does allow large problems to be run in a memory-limited environment, achieving scalability may be difficult due to inefficiencies in the current implementation of RMA operations.

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

    CERN Document Server

    Wilson, T; Carminati, F; Brun, R; Ferrari, A; Sala, P; Empl, A; MacGibbon, J

    2001-01-01

    We report on the progress of a current study aimed at developing a state-of-the-art Monte-Carlo computer simulation of the space radiation environment using advanced computer software techniques recently available at CERN, the European Laboratory for Particle Physics in Geneva, Switzerland. By taking the next-generation computer software appearing at CERN and adapting it to known problems in the implementation of space exploration strategies, this research is identifying changes necessary to bring these two advanced technologies together. The radiation transport tool being developed is tailored to the problem of taking measured space radiation fluxes impinging on the geometry of any particular spacecraft or planetary habitat and simulating the evolution of that flux through an accurate model of the spacecraft material. The simulation uses the latest known results in low-energy and high-energy physics. The output is a prediction of the detailed nature of the radiation environment experienced in space as well a...

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

    Science.gov (United States)

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

    2013-08-21

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

  1. Development of parallel monte carlo electron and photon transport (PMCEPT) code III: Applications to medical radiation physics

    Science.gov (United States)

    Kum, Oyeon; Han, Youngyih; Jeong, Hae Sun

    2012-05-01

    Minimizing the differences between dose distributions calculated at the treatment planning stage and those delivered to the patient is an essential requirement for successful radiotheraphy. Accurate calculation of dose distributions in the treatment planning process is important and can be done only by using a Monte Carlo calculation of particle transport. In this paper, we perform a further validation of our previously developed parallel Monte Carlo electron and photon transport (PMCEPT) code [Kum and Lee, J. Korean Phys. Soc. 47, 716 (2005) and Kim and Kum, J. Korean Phys. Soc. 49, 1640 (2006)] for applications to clinical radiation problems. A linear accelerator, Siemens' Primus 6 MV, was modeled and commissioned. A thorough validation includes both small fields, closely related to the intensity modulated radiation treatment (IMRT), and large fields. Two-dimensional comparisons with film measurements were also performed. The PMCEPT results, in general, agreed well with the measured data within a maximum error of about 2%. However, considering the experimental errors, the PMCEPT results can provide the gold standard of dose distributions for radiotherapy. The computing time was also much faster, compared to that needed for experiments, although it is still a bottleneck for direct applications to the daily routine treatment planning procedure.

  2. Tripoli-3: monte Carlo transport code for neutral particles - version 3.5 - users manual; Tripoli-3: code de transport des particules neutres par la methode de monte carlo - version 3.5 - manuel d'utilisation

    Energy Technology Data Exchange (ETDEWEB)

    Vergnaud, Th.; Nimal, J.C.; Chiron, M

    2001-07-01

    The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)

  3. MC21 v.6.0 - A Continuous-Energy Monte Carlo Particle Transport Code with Integrated Reactor Feedback Capabilities

    Science.gov (United States)

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

    2014-06-01

    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

  4. Mercury + VisIt: Integration of a Real-Time Graphical Analysis Capability into a Monte Carlo Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    O' Brien, M J; Procassini, R J; Joy, K I

    2009-03-09

    Validation of the problem definition and analysis of the results (tallies) produced during a Monte Carlo particle transport calculation can be a complicated, time-intensive processes. The time required for a person to create an accurate, validated combinatorial geometry (CG) or mesh-based representation of a complex problem, free of common errors such as gaps and overlapping cells, can range from days to weeks. The ability to interrogate the internal structure of a complex, three-dimensional (3-D) geometry, prior to running the transport calculation, can improve the user's confidence in the validity of the problem definition. With regard to the analysis of results, the process of extracting tally data from printed tables within a file is laborious and not an intuitive approach to understanding the results. The ability to display tally information overlaid on top of the problem geometry can decrease the time required for analysis and increase the user's understanding of the results. To this end, our team has integrated VisIt, a parallel, production-quality visualization and data analysis tool into Mercury, a massively-parallel Monte Carlo particle transport code. VisIt provides an API for real time visualization of a simulation as it is running. The user may select which plots to display from the VisIt GUI, or by sending VisIt a Python script from Mercury. The frequency at which plots are updated can be set and the user can visualize the simulation results as it is running.

  5. Comparison of dose estimates using the buildup-factor method and a Baryon transport code (BRYNTRN) with Monte Carlo results

    Science.gov (United States)

    Shinn, Judy L.; Wilson, John W.; Nealy, John E.; Cucinotta, Francis A.

    1990-01-01

    Continuing efforts toward validating the buildup factor method and the BRYNTRN code, which use the deterministic approach in solving radiation transport problems and are the candidate engineering tools in space radiation shielding analyses, are presented. A simplified theory of proton buildup factors assuming no neutron coupling is derived to verify a previously chosen form for parameterizing the dose conversion factor that includes the secondary particle buildup effect. Estimates of dose in tissue made by the two deterministic approaches and the Monte Carlo method are intercompared for cases with various thicknesses of shields and various types of proton spectra. The results are found to be in reasonable agreement but with some overestimation by the buildup factor method when the effect of neutron production in the shield is significant. Future improvement to include neutron coupling in the buildup factor theory is suggested to alleviate this shortcoming. Impressive agreement for individual components of doses, such as those from the secondaries and heavy particle recoils, are obtained between BRYNTRN and Monte Carlo results.

  6. Proton Dose Assessment to the Human Eye Using Monte Carlo N-Particle Transport Code (MCNPX)

    Science.gov (United States)

    2006-08-01

    objective of this project was to develop a simple MCNPX model of the human eye to approximate dose delivered from proton therapy. The calculated dose...computer code MCNPX that approximates dose delivered during proton therapy. The calculations considered proton interactions and secondary interactions...Volume Calculation The MCNPX code has limited ability to compute the volumes of defined cells. The dosimetric volumes in the outer wall of the eye are

  7. 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

    CERN Document Server

    White, M 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 tran...

  8. 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

  9. TARTNP: a coupled neutron--photon Monte Carlo transport code. [10-/sup 9/ to 20 MeV; in LLL FORTRAN

    Energy Technology Data Exchange (ETDEWEB)

    Plechaty, E.F.; Kimlinger, J.R.

    1976-07-04

    A Monte Carlo code was written that calculates the transport of neutrons, photons, and neutron-induced photons. The cross sections of these particles are derived from TARTNP's data base, the Evaluated Nuclear Data Library. The energy range of the neutron data in the Library is 10/sup -9/ MeV to 20 MeV; the photon energy range is 1 keV to 20 MeV. One of the chief advantages of the code is its flexibility: it allows up to 17 different kinds of output to be evaluated in the same problem.

  10. 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.

  11. 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)

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

    Science.gov (United States)

    Mosleh-Shirazi, Mohammad Amin; Zarrini-Monfared, Zinat; Karbasi, Sareh; 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 detector optimization.

  13. An Overview of the Monte Carlo Methods, Codes, & Applications Group

    Energy Technology Data Exchange (ETDEWEB)

    Trahan, Travis John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-08-30

    This report sketches the work of the Group to deliver first-principle Monte Carlo methods, production quality codes, and radiation transport-based computational and experimental assessments using the codes MCNP and MCATK for such applications as criticality safety, non-proliferation, nuclear energy, nuclear threat reduction and response, radiation detection and measurement, radiation health protection, and stockpile stewardship.

  14. Monte Carlo simulation code modernization

    CERN Document Server

    CERN. Geneva

    2015-01-01

    The continual development of sophisticated transport simulation algorithms allows increasingly accurate description of the effect of the passage of particles through matter. This modelling capability finds applications in a large spectrum of fields from medicine to astrophysics, and of course HEP. These new capabilities however come at the cost of a greater computational intensity of the new models, which has the effect of increasing the demands of computing resources. This is particularly true for HEP, where the demand for more simulation are driven by the need of both more accuracy and more precision, i.e. better models and more events. Usually HEP has relied on the "Moore's law" evolution, but since almost ten years the increase in clock speed has withered and computing capacity comes in the form of hardware architectures of many-core or accelerated processors. To harness these opportunities we need to adapt our code to concurrent programming models taking advantages of both SIMD and SIMT architectures. Th...

  15. Challenges of Monte Carlo Transport

    Energy Technology Data Exchange (ETDEWEB)

    Long, Alex Roberts [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-06-10

    These are slides from a presentation for Parallel Summer School at Los Alamos National Laboratory. Solving discretized partial differential equations (PDEs) of interest can require a large number of computations. We can identify concurrency to allow parallel solution of discrete PDEs. Simulated particles histories can be used to solve the Boltzmann transport equation. Particle histories are independent in neutral particle transport, making them amenable to parallel computation. Physical parameters and method type determine the data dependencies of particle histories. Data requirements shape parallel algorithms for Monte Carlo. Then, Parallel Computational Physics and Parallel Monte Carlo are discussed and, finally, the results are given. The mesh passing method greatly simplifies the IMC implementation and allows simple load-balancing. Using MPI windows and passive, one-sided RMA further simplifies the implementation by removing target synchronization. The author is very interested in implementations of PGAS that may allow further optimization for one-sided, read-only memory access (e.g. Open SHMEM). The MPICH_RMA_OVER_DMAPP option and library is required to make one-sided messaging scale on Trinitite - Moonlight scales poorly. Interconnect specific libraries or functions are likely necessary to ensure performance. BRANSON has been used to directly compare the current standard method to a proposed method on idealized problems. The mesh passing algorithm performs well on problems that are designed to show the scalability of the particle passing method. BRANSON can now run load-imbalanced, dynamic problems. Potential avenues of improvement in the mesh passing algorithm will be implemented and explored. A suite of test problems that stress DD methods will elucidate a possible path forward for production codes.

  16. Comparison of a 3-D multi-group SN particle transport code with Monte Carlo for intracavitary brachytherapy of the cervix uteri.

    Science.gov (United States)

    Gifford, Kent A; Wareing, Todd A; Failla, Gregory; Horton, John L; Eifel, Patricia J; Mourtada, Firas

    2009-12-03

    A patient dose distribution was calculated by a 3D multi-group S N particle transport code for intracavitary brachytherapy of the cervix uteri and compared to previously published Monte Carlo results. A Cs-137 LDR intracavitary brachytherapy CT data set was chosen from our clinical database. MCNPX version 2.5.c, was used to calculate the dose distribution. A 3D multi-group S N particle transport code, Attila version 6.1.1 was used to simulate the same patient. Each patient applicator was built in SolidWorks, a mechanical design package, and then assembled with a coordinate transformation and rotation for the patient. The SolidWorks exported applicator geometry was imported into Attila for calculation. Dose matrices were overlaid on the patient CT data set. Dose volume histograms and point doses were compared. The MCNPX calculation required 14.8 hours, whereas the Attila calculation required 22.2 minutes on a 1.8 GHz AMD Opteron CPU. Agreement between Attila and MCNPX dose calculations at the ICRU 38 points was within +/- 3%. Calculated doses to the 2 cc and 5 cc volumes of highest dose differed by not more than +/- 1.1% between the two codes. Dose and DVH overlays agreed well qualitatively. Attila can calculate dose accurately and efficiently for this Cs-137 CT-based patient geometry. Our data showed that a three-group cross-section set is adequate for Cs-137 computations. Future work is aimed at implementing an optimized version of Attila for radiotherapy calculations.

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

    Science.gov (United States)

    Iwamoto, Yosuke; Ogawa, Tatsuhiko

    2017-04-01

    Because primary knock-on atoms (PKAs) create point defects and clusters in materials that are irradiated with neutrons, it is important to validate the calculations of recoil cross section spectra that are used to estimate radiation damage in materials. Here, the recoil cross section spectra of fission- and fusion-relevant materials were calculated using the Event Generator Mode (EGM) of the Particle and Heavy Ion Transport code System (PHITS) and also using the data processing code NJOY2012 with the nuclear data libraries TENDL2015, ENDF/BVII.1, and JEFF3.2. The heating number, which is the integral of the recoil cross section spectra, was also calculated using PHITS-EGM and compared with data extracted from the ACE files of TENDL2015, ENDF/BVII.1, and JENDL4.0. In general, only a small difference was found between the PKA spectra of PHITS + TENDL2015 and NJOY + TENDL2015. From analyzing the recoil cross section spectra extracted from the nuclear data libraries using NJOY2012, we found that the recoil cross section spectra were incorrect for 72Ge, 75As, 89Y, and 109Ag in the ENDF/B-VII.1 library, and for 90Zr and 55Mn in the JEFF3.2 library. From analyzing the heating number, we found that the data extracted from the ACE file of TENDL2015 for all nuclides were problematic in the neutron capture region because of incorrect data regarding the emitted gamma energy. However, PHITS + TENDL2015 can calculate PKA spectra and heating numbers correctly.

  18. Production of energetic light fragments in extensions of the CEM and LAQGSM event generators of the Monte Carlo transport code MCNP6

    Science.gov (United States)

    Mashnik, Stepan G.; Kerby, Leslie M.; Gudima, Konstantin K.; Sierk, Arnold J.; Bull, Jeffrey S.; James, Michael R.

    2017-03-01

    We extend the cascade-exciton model (CEM), and the Los Alamos version of the quark-gluon string model (LAQGSM), event generators of the Monte Carlo N -particle transport code version 6 (MCNP6), to describe production of energetic light fragments (LF) heavier than 4He from various nuclear reactions induced by particles and nuclei at energies up to about 1 TeV/nucleon. In these models, energetic LF can be produced via Fermi breakup, preequilibrium emission, and coalescence of cascade particles. Initially, we study several variations of the Fermi breakup model and choose the best option for these models. Then, we extend the modified exciton model (MEM) used by these codes to account for a possibility of multiple emission of up to 66 types of particles and LF (up to 28Mg) at the preequilibrium stage of reactions. Then, we expand the coalescence model to allow coalescence of LF from nucleons emitted at the intranuclear cascade stage of reactions and from lighter clusters, up to fragments with mass numbers A ≤7 , in the case of CEM, and A ≤12 , in the case of LAQGSM. Next, we modify MCNP6 to allow calculating and outputting spectra of LF and heavier products with arbitrary mass and charge numbers. The improved version of CEM is implemented into MCNP6. Finally, we test the improved versions of CEM, LAQGSM, and MCNP6 on a variety of measured nuclear reactions. The modified codes give an improved description of energetic LF from particle- and nucleus-induced reactions; showing a good agreement with a variety of available experimental data. They have an improved predictive power compared to the previous versions and can be used as reliable tools in simulating applications involving such types of reactions.

  19. NASA Space Radiation Transport Code Development Consortium.

    Science.gov (United States)

    Townsend, Lawrence W

    2005-01-01

    Recently, NASA established a consortium involving the University of Tennessee (lead institution), the University of Houston, Roanoke College and various government and national laboratories, to accelerate the development of a standard set of radiation transport computer codes for NASA human exploration applications. This effort involves further improvements of the Monte Carlo codes HETC and FLUKA and the deterministic code HZETRN, including developing nuclear reaction databases necessary to extend the Monte Carlo codes to carry out heavy ion transport, and extending HZETRN to three dimensions. The improved codes will be validated by comparing predictions with measured laboratory transport data, provided by an experimental measurements consortium, and measurements in the upper atmosphere on the balloon-borne Deep Space Test Bed (DSTB). In this paper, we present an overview of the consortium members and the current status and future plans of consortium efforts to meet the research goals and objectives of this extensive undertaking.

  20. Implict Monte Carlo Radiation Transport Simulations of Four Test Problems

    Energy Technology Data Exchange (ETDEWEB)

    Gentile, N

    2007-08-01

    Radiation transport codes, like almost all codes, are difficult to develop and debug. It is helpful to have small, easy to run test problems with known answers to use in development and debugging. It is also prudent to re-run test problems periodically during development to ensure that previous code capabilities have not been lost. We describe four radiation transport test problems with analytic or approximate analytic answers. These test problems are suitable for use in debugging and testing radiation transport codes. We also give results of simulations of these test problems performed with an Implicit Monte Carlo photonics code.

  1. Verification of Monte Carlo transport codes against measured small angle p-, d-, and t-emission in carbon fragmentation at 600 MeV/nucleon

    Energy Technology Data Exchange (ETDEWEB)

    Abramov, B. M. [Inst. of Theoretical and Experimental Physics (ITEP), Moscow (Russian Federation); Alekseev, P. N. [Inst. of Theoretical and Experimental Physics (ITEP), Moscow (Russian Federation); Borodin, Yu. A. [Inst. of Theoretical and Experimental Physics (ITEP), Moscow (Russian Federation); Bulychjov, S. A. [Inst. of Theoretical and Experimental Physics (ITEP), Moscow (Russian Federation); Dukhovskoy, I. A. [Inst. of Theoretical and Experimental Physics (ITEP), Moscow (Russian Federation); Krutenkova, A. P. [Inst. of Theoretical and Experimental Physics (ITEP), Moscow (Russian Federation); Martemianov, M. A. [Inst. of Theoretical and Experimental Physics (ITEP), Moscow (Russian Federation); Matsyuk, M. A. [Inst. of Theoretical and Experimental Physics (ITEP), Moscow (Russian Federation); Turdakina, E. N. [Inst. of Theoretical and Experimental Physics (ITEP), Moscow (Russian Federation); Khanov, A. I. [Inst. of Theoretical and Experimental Physics (ITEP), Moscow (Russian Federation); Mashnik, Stepan Georgievich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-02-03

    Momentum spectra of hydrogen isotopes have been measured at 3.5° from 12C fragmentation on a Be target. Momentum spectra cover both the region of fragmentation maximum and the cumulative region. Differential cross sections span five orders of magnitude. The data are compared to predictions of four Monte Carlo codes: QMD, LAQGSM, BC, and INCL++. There are large differences between the data and predictions of some models in the high momentum region. The INCL++ code gives the best and almost perfect description of the data.

  2. PARALLELIZATION AND PERFECTION OF MCNP MONTE CARLO PARTICLE TRANSPORT CODE IN MPI%粒子输运蒙特卡罗程序MCNP在MPI下的并行化及完善

    Institute of Scientific and Technical Information of China (English)

    邓力; 刘杰; 张文勇

    2003-01-01

    The particle transport Monte Carlo code MCNP had been realized the paral-lelization in MPI (Message Passing Interface) in 1999. But due to adopting the leap random number producer, some differences were existed between the parallel result and the serial result. Now the same results have been achieved by using the segment random number. The speedup of the applied problem is the liner ups to 53 in 64-Processors and the parallel efficiencv is up to 83% in 64-Processors.

  3. Effect of the electron transport through thin slabs on the simulation of linear electron accelerators of use in therapy: A comparative study of various Monte Carlo codes

    Energy Technology Data Exchange (ETDEWEB)

    Vilches, M. [Servicio de Fisica y Proteccion Radiologica, Hospital Regional Universitario ' Virgen de las Nieves' , Avda. de las Fuerzas Armadas, 2, E-18014 Granada (Spain)], E-mail: mvilches@ugr.es; Garcia-Pareja, S. [Servicio de Radiofisica Hospitalaria, Hospital Regional Universitario ' Carlos Haya' , Avda. Carlos Haya, s/n, E-29010 Malaga (Spain); Guerrero, R. [Servicio de Radiofisica, Hospital Universitario ' San Cecilio' , Avda. Dr. Oloriz, 16, E-18012 Granada (Spain); Anguiano, M.; Lallena, A.M. [Departamento de Fisica Atomica, Molecular y Nuclear, Universidad de Granada, E-18071 Granada (Spain)

    2007-09-21

    When a therapeutic electron linear accelerator is simulated using a Monte Carlo (MC) code, the tuning of the initial spectra and the renormalization of dose (e.g., to maximum axial dose) constitute a common practice. As a result, very similar depth dose curves are obtained for different MC codes. However, if renormalization is turned off, the results obtained with the various codes disagree noticeably. The aim of this work is to investigate in detail the reasons of this disagreement. We have found that the observed differences are due to non-negligible differences in the angular scattering of the electron beam in very thin slabs of dense material (primary foil) and thick slabs of very low density material (air). To gain insight, the effects of the angular scattering models considered in various MC codes on the dose distribution in a water phantom are discussed using very simple geometrical configurations for the LINAC. The MC codes PENELOPE 2003, PENELOPE 2005, GEANT4, GEANT3, EGSnrc and MCNPX have been used.

  4. Proton therapy Monte Carlo SRNA-VOX code

    Directory of Open Access Journals (Sweden)

    Ilić Radovan D.

    2012-01-01

    Full Text Available The most powerful feature of the Monte Carlo method is the possibility of simulating all individual particle interactions in three dimensions and performing numerical experiments with a preset error. These facts were the motivation behind the development of a general-purpose Monte Carlo SRNA program for proton transport simulation in technical systems described by standard geometrical forms (plane, sphere, cone, cylinder, cube. Some of the possible applications of the SRNA program are: (a a general code for proton transport modeling, (b design of accelerator-driven systems, (c simulation of proton scattering and degrading shapes and composition, (d research on proton detectors; and (e radiation protection at accelerator installations. This wide range of possible applications of the program demands the development of various versions of SRNA-VOX codes for proton transport modeling in voxelized geometries and has, finally, resulted in the ISTAR package for the calculation of deposited energy distribution in patients on the basis of CT data in radiotherapy. All of the said codes are capable of using 3-D proton sources with an arbitrary energy spectrum in an interval of 100 keV to 250 MeV.

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

    Energy Technology Data Exchange (ETDEWEB)

    Cupini, E. [ENEA, Centro Ricerche `Ezio Clementel`, Bologna (Italy). Dipt. Innovazione; Borgia, M.G. [ENEA, Centro Ricerche `Ezio Clementel`, Bologna (Italy). Dipt. Energia; Premuda, M. [Consiglio Nazionale delle Ricerche, Bologna (Italy). Ist. FISBAT

    1997-03-01

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

  6. Applications guide to the MORSE Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Cramer, S.N.

    1985-08-01

    A practical guide for the implementation of the MORESE-CG Monte Carlo radiation transport computer code system is presented. The various versions of the MORSE code are compared and contrasted, and the many references dealing explicitly with the MORSE-CG code are reviewed. The treatment of angular scattering is discussed, and procedures for obtaining increased differentiality of results in terms of reaction types and nuclides from a multigroup Monte Carlo code are explained in terms of cross-section and geometry data manipulation. Examples of standard cross-section data input and output are shown. Many other features of the code system are also reviewed, including (1) the concept of primary and secondary particles, (2) fission neutron generation, (3) albedo data capability, (4) DOMINO coupling, (5) history file use for post-processing of results, (6) adjoint mode operation, (7) variance reduction, and (8) input/output. In addition, examples of the combinatorial geometry are given, and the new array of arrays geometry feature (MARS) and its three-dimensional plotting code (JUNEBUG) are presented. Realistic examples of user routines for source, estimation, path-length stretching, and cross-section data manipulation are given. A deatiled explanation of the coupling between the random walk and estimation procedure is given in terms of both code parameters and physical analogies. The operation of the code in the adjoint mode is covered extensively. The basic concepts of adjoint theory and dimensionality are discussed and examples of adjoint source and estimator user routines are given for all common situations. Adjoint source normalization is explained, a few sample problems are given, and the concept of obtaining forward differential results from adjoint calculations is covered. Finally, the documentation of the standard MORSE-CG sample problem package is reviewed and on-going and future work is discussed.

  7. Recent developments in the Los Alamos radiation transport code system

    Energy Technology Data Exchange (ETDEWEB)

    Forster, R.A.; Parsons, K. [Los Alamos National Lab., NM (United States)

    1997-06-01

    A brief progress report on updates to the Los Alamos Radiation Transport Code System (LARTCS) for solving criticality and fixed-source problems is provided. LARTCS integrates the Diffusion Accelerated Neutral Transport (DANT) discrete ordinates codes with the Monte Carlo N-Particle (MCNP) code. The LARCTS code is being developed with a graphical user interface for problem setup and analysis. Progress in the DANT system for criticality applications include a two-dimensional module which can be linked to a mesh-generation code and a faster iteration scheme. Updates to MCNP Version 4A allow statistical checks of calculated Monte Carlo results.

  8. Monte Carlo methods for particle transport

    CERN Document Server

    Haghighat, Alireza

    2015-01-01

    The Monte Carlo method has become the de facto standard in radiation transport. Although powerful, if not understood and used appropriately, the method can give misleading results. Monte Carlo Methods for Particle Transport teaches appropriate use of the Monte Carlo method, explaining the method's fundamental concepts as well as its limitations. Concise yet comprehensive, this well-organized text: * Introduces the particle importance equation and its use for variance reduction * Describes general and particle-transport-specific variance reduction techniques * Presents particle transport eigenvalue issues and methodologies to address these issues * Explores advanced formulations based on the author's research activities * Discusses parallel processing concepts and factors affecting parallel performance Featuring illustrative examples, mathematical derivations, computer algorithms, and homework problems, Monte Carlo Methods for Particle Transport provides nuclear engineers and scientists with a practical guide ...

  9. 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.

  10. Monte Carlo simulation of nuclear energy study (II). Annual report on Nuclear Code Evaluation Committee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-01-01

    In the report, research results discussed in 1999 fiscal year at Nuclear Code Evaluation Committee of Nuclear Code Research Committee were summarized. Present status of Monte Carlo simulation on nuclear energy study was described. Especially, besides of criticality, shielding and core analyses, present status of applications to risk and radiation damage analyses, high energy transport and nuclear theory calculations of Monte Carlo Method was described. The 18 papers are indexed individually. (J.P.N.)

  11. Sub-Transport Layer Coding

    DEFF Research Database (Denmark)

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

    2014-01-01

    Packet losses in wireless networks dramatically curbs the performance of TCP. This paper introduces a simple coding shim that aids IP-layer traffic in lossy environments while being transparent to transport layer protocols. The proposed coding approach enables erasure correction while being...... oblivious to the congestion control algorithms of the utilised transport layer protocol. Although our coding shim is indifferent towards the transport layer protocol, we focus on the performance of TCP when ran on top of our proposed coding mechanism due to its widespread use. The coding shim provides gains...

  12. 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.

  13. FREYA-a new Monte Carlo code for improved modeling of fission chains

    Energy Technology Data Exchange (ETDEWEB)

    Hagmann, C A; Randrup, J; Vogt, R L

    2012-06-12

    A new simulation capability for modeling of individual fission events and chains and the transport of fission products in materials is presented. FREYA ( Fission Yield Event Yield Algorithm ) is a Monte Carlo code for generating fission events providing correlated kinematic information for prompt neutrons, gammas, and fragments. As a standalone code, FREYA calculates quantities such as multiplicity-energy, angular, and gamma-neutron energy sharing correlations. To study materials with multiplication, shielding effects, and detectors, we have integrated FREYA into the general purpose Monte Carlo code MCNP. This new tool will allow more accurate modeling of detector responses including correlations and the development of SNM detectors with increased sensitivity.

  14. The impact of advances in computer technology on particle transport Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Martin, W.R. [Michigan Univ., Ann Arbor, MI (United States). Dept. of Nuclear Engineering; Rathkopf, J.A. [Lawrence Livermore National Lab., CA (United States); Brown, F.B. [Knolls Atomic Power Lab., Schenectady, NY (United States)

    1992-01-21

    Advances in computer technology, including hardware, architectural, and software advances, have led to dramatic gains in computer performance over the past decade. We summarize these performance trends and discuss the extent to which particle transport Monte Carlo codes have been able to take advantage of these performance gains. We consider MIMD, SIMD, and parallel distributed computer configurations for particle transport Monte Carlo applications. Some specific experience with vectorization and parallelization of production Monte Carlo codes is included. The topic of parallel random number generation is discussed in some detail. Finally, some software issues that hinder the implementation of Monte Carlo methods on parallel processors are addressed.

  15. Fluence to absorbed dose, effective dose and gray equivalent conversion coefficients for iron nuclei from 10 MeV to 1 TeV, calculated using Monte Carlo radiation transport code MCNPX 2.7.A.

    Science.gov (United States)

    Copeland, Kyle; Parker, Donald E; Friedberg, Wallace

    2010-03-01

    Conversion coefficients have been calculated for fluence-to-absorbed dose, fluence-to-effective dose and fluence-to-gray equivalent for isotropic exposure of an adult male and an adult female to (56)Fe(26+) in the energy range of 10 MeV to 1 TeV (0.01-1000 GeV). The coefficients were calculated using Monte Carlo transport code MCNPX 2.7.A and BodyBuilder 1.3 anthropomorphic phantoms modified to allow calculation of effective dose using tissues and tissue weighting factors from either the 1990 or 2007 recommendations of the International Commission on Radiological Protection (ICRP) and gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. Calculations using ICRP 2007 recommendations result in fluence-to-effective dose conversion coefficients that are almost identical at most energies to those calculated using ICRP 1990 recommendations.

  16. Alpha particles at energies of 10 MeV to 1 TeV: conversion coefficients for fluence-to-absorbed dose, effective dose, and gray equivalent, calculated using Monte Carlo radiation transport code MCNPX 2.7.A.

    Science.gov (United States)

    Copeland, Kyle; Parker, Donald E; Friedberg, Wallace

    2010-03-01

    Conversion coefficients have been calculated for fluence to absorbed dose, fluence to effective dose and fluence to gray equivalent, for isotropic exposure to alpha particles in the energy range of 10 MeV to 1 TeV (0.01-1000 GeV). The coefficients were calculated using Monte Carlo transport code MCNPX 2.7.A and BodyBuilder 1.3 anthropomorphic phantoms modified to allow calculation of effective dose to a Reference Person using tissues and tissue weighting factors from 1990 and 2007 recommendations of the International Commission on Radiological Protection (ICRP) and gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. Coefficients for effective dose are within 30 % of those calculated using ICRP 1990 recommendations.

  17. A Deterministic Transport Code for Space Environment Electrons

    Science.gov (United States)

    Nealy, John E.; Chang, C. K.; Norman, Ryan B.; Blattnig, Steve R.; Badavi, Francis F.; Adamczyk, Anne M.

    2010-01-01

    A deterministic computational procedure has been developed to describe transport of space environment electrons in various shield media. This code is an upgrade and extension of an earlier electron code. Whereas the former code was formulated on the basis of parametric functions derived from limited laboratory data, the present code utilizes well established theoretical representations to describe the relevant interactions and transport processes. The shield material specification has been made more general, as have the pertinent cross sections. A combined mean free path and average trajectory approach has been used in the transport formalism. Comparisons with Monte Carlo calculations are presented.

  18. 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.

  19. Criticality benchmarks validation of the Monte Carlo code TRIPOLI-2

    Energy Technology Data Exchange (ETDEWEB)

    Maubert, L. (Commissariat a l' Energie Atomique, Inst. de Protection et de Surete Nucleaire, Service d' Etudes de Criticite, 92 - Fontenay-aux-Roses (France)); Nouri, A. (Commissariat a l' Energie Atomique, Inst. de Protection et de Surete Nucleaire, Service d' Etudes de Criticite, 92 - Fontenay-aux-Roses (France)); Vergnaud, T. (Commissariat a l' Energie Atomique, Direction des Reacteurs Nucleaires, Service d' Etudes des Reacteurs et de Mathematique Appliquees, 91 - Gif-sur-Yvette (France))

    1993-04-01

    The three-dimensional energy pointwise Monte-Carlo code TRIPOLI-2 includes metallic spheres of uranium and plutonium, nitrate plutonium solutions, square and triangular pitch assemblies of uranium oxide. Results show good agreements between experiments and calculations, and avoid a part of the code and its ENDF-B4 library validation. (orig./DG)

  20. Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations

    Energy Technology Data Exchange (ETDEWEB)

    Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    A burn-up calculation of VVER's cores by Monte-Carlo code is complex process and requires large computational costs. This fact makes Monte-Carlo codes usage complicated for project and operating calculations. Previously prepared isotopic compositions are proposed to use for the Monte-Carlo code (MCU) calculations of different states of VVER's core with burnt fuel. Isotopic compositions are proposed to calculate by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by engineering codes (TVS-M, PERMAK-A). The multiplication factors and power distributions of FA and VVER with infinite height are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The MCU calculation data were compared with the data which were obtained by engineering codes.

  1. The Monte Carlo code MCSHAPE: Main features and recent developments

    Energy Technology Data Exchange (ETDEWEB)

    Scot, Viviana, E-mail: viviana.scot@unibo.it; Fernandez, Jorge E.

    2015-06-01

    MCSHAPE is a general purpose Monte Carlo code developed at the University of Bologna to simulate the diffusion of X- and gamma-ray photons with the special feature of describing the full evolution of the photon polarization state along the interactions with the target. The prevailing photon–matter interactions in the energy range 1–1000 keV, Compton and Rayleigh scattering and photoelectric effect, are considered. All the parameters that characterize the photon transport can be suitably defined: (i) the source intensity, (ii) its full polarization state as a function of energy, (iii) the number of collisions, and (iv) the energy interval and resolution of the simulation. It is possible to visualize the results for selected groups of interactions. MCSHAPE simulates the propagation in heterogeneous media of polarized photons (from synchrotron sources) or of partially polarized sources (from X-ray tubes). In this paper, the main features of MCSHAPE are illustrated with some examples and a comparison with experimental data. - Highlights: • MCSHAPE is an MC code for the simulation of the diffusion of photons in the matter. • It includes the proper description of the evolution of the photon polarization state. • The polarization state is described by means of the Stokes vector, I, Q, U, V. • MCSHAPE includes the computation of the detector influence in the measured spectrum. • MCSHAPE features are illustrated with examples and comparison with experiments.

  2. Monte Carlo radiation transport in external beam radiotherapy

    OpenAIRE

    Çeçen, Yiğit

    2013-01-01

    The use of Monte Carlo in radiation transport is an effective way to predict absorbed dose distributions. Monte Carlo modeling has contributed to a better understanding of photon and electron transport by radiotherapy physicists. The aim of this review is to introduce Monte Carlo as a powerful radiation transport tool. In this review, photon and electron transport algorithms for Monte Carlo techniques are investigated and a clinical linear accelerator model is studied for external beam radiot...

  3. MCOR - Monte Carlo depletion code for reference LWR calculations

    Energy Technology Data Exchange (ETDEWEB)

    Puente Espel, Federico, E-mail: fup104@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Tippayakul, Chanatip, E-mail: cut110@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Ivanov, Kostadin, E-mail: kni1@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Misu, Stefan, E-mail: Stefan.Misu@areva.com [AREVA, AREVA NP GmbH, Erlangen (Germany)

    2011-04-15

    Research highlights: > Introduction of a reference Monte Carlo based depletion code with extended capabilities. > Verification and validation results for MCOR. > Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations. Additionally

  4. A Monte Carlo simulation code for calculating damage and particle transport in solids: The case for electron-bombarded solids for electron energies up to 900 MeV

    Science.gov (United States)

    Yan, Qiang; Shao, Lin

    2017-03-01

    Current popular Monte Carlo simulation codes for simulating electron bombardment in solids focus primarily on electron trajectories, instead of electron-induced displacements. Here we report a Monte Carol simulation code, DEEPER (damage creation and particle transport in matter), developed for calculating 3-D distributions of displacements produced by electrons of incident energies up to 900 MeV. Electron elastic scattering is calculated by using full-Mott cross sections for high accuracy, and primary-knock-on-atoms (PKAs)-induced damage cascades are modeled using ZBL potential. We compare and show large differences in 3-D distributions of displacements and electrons in electron-irradiated Fe. The distributions of total displacements are similar to that of PKAs at low electron energies. But they are substantially different for higher energy electrons due to the shifting of PKA energy spectra towards higher energies. The study is important to evaluate electron-induced radiation damage, for the applications using high flux electron beams to intentionally introduce defects and using an electron analysis beam for microstructural characterization of nuclear materials.

  5. Parallelization of Monte Carlo codes MVP/GMVP

    Energy Technology Data Exchange (ETDEWEB)

    Nagaya, Yasunobu; Mori, Takamasa; Nakagawa, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Sasaki, Makoto

    1998-03-01

    General-purpose Monte Carlo codes MVP/GMVP are well-vectorized and thus enable us to perform high-speed Monte Carlo calculations. In order to achieve more speedups, we parallelized the codes on the different types of the parallel processing platforms. The platforms reported are a distributed-memory vector-parallel computer Fujitsu VPP500, a distributed-memory massively parallel computer Intel Paragon and a distributed-memory scalar-parallel computer Hitachi SR2201. As mentioned generally, ideal speedup could be obtained for large-scale problems but parallelization efficiency got worse as the batch size per a processing element (PE) was smaller. (author)

  6. A Monte Carlo code for ion beam therapy

    CERN Multimedia

    Anaïs Schaeffer

    2012-01-01

    Initially developed for applications in detector and accelerator physics, the modern Fluka Monte Carlo code is now used in many different areas of nuclear science. Over the last 25 years, the code has evolved to include new features, such as ion beam simulations. Given the growing use of these beams in cancer treatment, Fluka simulations are being used to design treatment plans in several hadron-therapy centres in Europe.   Fluka calculates the dose distribution for a patient treated at CNAO with proton beams. The colour-bar displays the normalized dose values. Fluka is a Monte Carlo code that very accurately simulates electromagnetic and nuclear interactions in matter. In the 1990s, in collaboration with NASA, the code was developed to predict potential radiation hazards received by space crews during possible future trips to Mars. Over the years, it has become the standard tool to investigate beam-machine interactions, radiation damage and radioprotection issues in the CERN accelerator com...

  7. 多群蒙卡输运与点燃耗耦合程序系统TRITON基准验证%Benchmark Verification of Multi-group Monte Carlo Transport and Point-Burnup Codes Coupling System TRITON

    Institute of Scientific and Technical Information of China (English)

    武祥; 若夕子; 于涛; 谢金森; 陈昊威

    2014-01-01

    TRITON couples multi group Monte Carlo Transport code KENO V. a and point-burnup code ORIGEN-S. It features adaptability on complex geometries,flexible processing ability on cross section and rapid calculating speed. Based on the thorium-based fuel cell benchmark of Idaho National Laboratory ( INL) ,the verification on TRITON burnup calcu-lation was performed,which showed good coincidence with the result of MOCUP code by INL. Furthermore, the results of burnup isotopes selection schemes in TRITON showed that,for thorium based fuel,only important nuclides on Th-U cycle was included,correct results can be obtained by TRITON. Conclusions in the present paper will support further applications of TRITON.%TRITON程序系统耦合了多群蒙特卡罗输运程序KENO V. a与点燃耗程序ORIGEN-S,具有几何适应性强、截面处理能力灵活、计算速度快等显著特点.本文基于爱达荷国家实验室( INL)钍基燃料元件燃耗基准题,开展了TRITON程序燃耗功能的验证,结果与INL采用MOCUP程序给出的结果吻合很好.同时,燃耗核素选取对TRITON计算结果的影响分析表明对于钍基燃料,只有在考虑Th-U循环重要核素的前提下,TRITON才能给出正确结果.上述结论为TRITON程序的应用奠定了基础.

  8. Discrete angle biasing in Monte Carlo radiation transport

    Energy Technology Data Exchange (ETDEWEB)

    Cramer, S.N.

    1988-05-01

    An angular biasing procedure is presented for use in Monte Carlo radiation transport with discretized scattering angle data. As in more general studies, the method is shown to reduce statistical weight fluctuations when it is combined with the exponential transformation. This discrete data application has a simple analytic form which is problem independent. The results from a sample problem illustrate the variance reduction and efficiency characteristics of the combined biasing procedures, and a large neutron and gamma ray integral experiment is also calculated. A proposal is given for the possible code generation of the biasing parameter p and the preferential direction /ovr/Omega///sub 0/ used in the combined biasing schemes.

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

    Science.gov (United States)

    Copeland, Kyle; Parker, Donald E; Friedberg, Wallace

    2011-01-01

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

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

    Science.gov (United States)

    Copeland, Kyle; Parker, Donald E; Friedberg, Wallace

    2010-12-01

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

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

    Science.gov (United States)

    Copeland, Kyle; Parker, Donald E; Friedberg, Wallace

    2010-12-01

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

  12. 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.

  13. On the use of SERPENT Monte Carlo code to generate few group diffusion constants

    Energy Technology Data Exchange (ETDEWEB)

    Piovezan, Pamela, E-mail: pamela.piovezan@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Carluccio, Thiago; Domingos, Douglas Borges; Rossi, Pedro Russo; Mura, Luiz Felipe, E-mail: fermium@cietec.org.b, E-mail: thiagoc@ipen.b [Fermium Tecnologia Nuclear, Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The accuracy of diffusion reactor codes strongly depends on the quality of the groups constants processing. For many years, the generation of such constants was based on 1-D infinity cell transport calculations. Some developments using collision probability or the method of characteristics allow, nowadays, 2-D assembly group constants calculations. However, these 1-D and 2-D codes how some limitations as , for example, on complex geometries and in the neighborhood of heavy absorbers. On the other hand, since Monte Carlos (MC) codes provide accurate neutro flux distributions, the possibility of using these solutions to provide group constants to full-core reactor diffusion simulators has been recently investigated, especially for the cases in which the geometry and reactor types are beyond the capability of the conventional deterministic lattice codes. The two greatest difficulties on the use of MC codes to group constant generation are the computational costs and the methodological incompatibility between analog MC particle transport simulation and deterministic transport methods based in several approximations. The SERPENT code is a 3-D continuous energy MC transport code with built-in burnup capability that was specially optimized to generate these group constants. In this work, we present the preliminary results of using the SERPENT MC code to generate 3-D two-group diffusion constants for a PWR like assembly. These constants were used in the CITATION diffusion code to investigate the effects of the MC group constants determination on the neutron multiplication factor diffusion estimate. (author)

  14. The Monte Carlo estimation of an effect of uncertainties in initial data on solving the transport equation by means of the MCU code

    Science.gov (United States)

    Oleynik, D. S.

    2015-12-01

    A new version of the tally module of the MCU software package is developed in which the approach for taking directly into account the uncertainty in initial data is implemented that is recommended by the international standard on estimating the uncertainty in results of measuring (ISO 13005). The new module makes it possible to evaluate the effect of uncertainty in initial data (caused by technological tolerances in fabrication of structural members of the core) on neutronic characteristics of the reactor. The developed software is adapted to parallel computing with the use of multiprocessor computers, which significantly reduces the computation time: the parallelization coefficient is almost equal to 1. Testing is performed by examples of solving the problem on criticality for the Godiva benchmark experiment and also for the infinite lattice of fuel assemblies of the VVER-440, VVER-1000, and VVER-1200. The results of calculations of the uncertainty in neutronic characteristics (effective multiplication factor, fission reaction rate), which is caused by uncertainties in initial data due to technological tolerances, are compared (in the first case) to the published results obtained using the precision MCNP5 code and (in the second case) to those obtained by means of the RADAR engineering program. A good agreement of results is achieved for all cases.

  15. The Monte Carlo estimation of an effect of uncertainties in initial data on solving the transport equation by means of the MCU code

    Energy Technology Data Exchange (ETDEWEB)

    Oleynik, D. S., E-mail: oleynik-ds@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    A new version of the tally module of the MCU software package is developed in which the approach for taking directly into account the uncertainty in initial data is implemented that is recommended by the international standard on estimating the uncertainty in results of measuring (ISO 13005). The new module makes it possible to evaluate the effect of uncertainty in initial data (caused by technological tolerances in fabrication of structural members of the core) on neutronic characteristics of the reactor. The developed software is adapted to parallel computing with the use of multiprocessor computers, which significantly reduces the computation time: the parallelization coefficient is almost equal to 1. Testing is performed by examples of solving the problem on criticality for the Godiva benchmark experiment and also for the infinite lattice of fuel assemblies of the VVER-440, VVER-1000, and VVER-1200. The results of calculations of the uncertainty in neutronic characteristics (effective multiplication factor, fission reaction rate), which is caused by uncertainties in initial data due to technological tolerances, are compared (in the first case) to the published results obtained using the precision MCNP5 code and (in the second case) to those obtained by means of the RADAR engineering program. A good agreement of results is achieved for all cases.

  16. A semianalytic Monte Carlo code for modelling LIDAR measurements

    Science.gov (United States)

    Palazzi, Elisa; Kostadinov, Ivan; Petritoli, Andrea; Ravegnani, Fabrizio; Bortoli, Daniele; Masieri, Samuele; Premuda, Margherita; Giovanelli, Giorgio

    2007-10-01

    LIDAR (LIght Detection and Ranging) is an optical active remote sensing technology with many applications in atmospheric physics. Modelling of LIDAR measurements appears useful approach for evaluating the effects of various environmental variables and scenarios as well as of different measurement geometries and instrumental characteristics. In this regard a Monte Carlo simulation model can provide a reliable answer to these important requirements. A semianalytic Monte Carlo code for modelling LIDAR measurements has been developed at ISAC-CNR. The backscattered laser signal detected by the LIDAR system is calculated in the code taking into account the contributions due to the main atmospheric molecular constituents and aerosol particles through processes of single and multiple scattering. The contributions by molecular absorption, ground and clouds reflection are evaluated too. The code can perform simulations of both monostatic and bistatic LIDAR systems. To enhance the efficiency of the Monte Carlo simulation, analytical estimates and expected value calculations are performed. Artificial devices (such as forced collision, local forced collision, splitting and russian roulette) are moreover foreseen by the code, which can enable the user to drastically reduce the variance of the calculation.

  17. 脉冲中子-裂变中子探测铀黄饼的MCNP模拟%The Monte Carlo N particle transport code simulation of pulsed neutron-fission neutron uranium yellowcake exploration

    Institute of Scientific and Technical Information of China (English)

    张坤明; 张雄杰; 瞿金辉; 汤彬

    2015-01-01

    利用MCNP程序模拟研究脉冲中子-裂变中子探测铀黄饼,采用脉冲式中子源,利用氦三管中子探测器记录裂变中子,得到铀黄饼中的铀含量信息。通过对14 MeV脉冲中子源和产生的裂变中子在不同铀含量模型中的输运计算,分析了裂变中子与铀含量的关系。结果表明:利用裂变超热中子衰减时间谱,可以确定铀黄饼中的铀含量;通过对热中子衰减时间谱进行校正,可以提高铀黄饼中铀含量计算结果的准确度。%The Monte Carlo N particle transport code ( MCNP ) is used to simulate how to explore the uranium yel⁃lowcake by using the pulsed neutron⁃fission neutron ( PNFN) method. In order to obtain uranium yellowcake quan⁃titation, pulsed neutron source was used, prompt fission neutrons were detected by using the neutron detector. Un⁃der the condition of different uranium quantitation models, the transport of the 14 MeV pulsed neutron source and the released fission neutron were calculated. On the basis of these, the relationship between fission neutron and ura⁃nium quantitation was studied. The results show that using the epithermal neutron time decay spectrum, the urani⁃um yellowcake quantitation can be determined; the precision of the uranium yellowcake quantitation could be in⁃creased by the correction of thermal neutron time decay spectrum.

  18. Domain Decomposition strategy for pin-wise full-core Monte Carlo depletion calculation with the reactor Monte Carlo Code

    Energy Technology Data Exchange (ETDEWEB)

    Liang, Jingang; Wang, Kan; Qiu, Yishu [Dept. of Engineering Physics, LiuQing Building, Tsinghua University, Beijing (China); Chai, Xiao Ming; Qiang, Sheng Long [Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu (China)

    2016-06-15

    Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC) codes in accomplishing pin-wise three-dimensional (3D) full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC) code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.

  19. Effects of physics change in Monte Carlo code on electron pencil beam dose distributions

    Energy Technology Data Exchange (ETDEWEB)

    Toutaoui, Abdelkader, E-mail: toutaoui.aek@gmail.com [Departement de Physique Medicale, Centre de Recherche Nucleaire d' Alger, 2 Bd Frantz Fanon BP399 Alger RP, Algiers (Algeria); Khelassi-Toutaoui, Nadia, E-mail: nadiakhelassi@yahoo.fr [Departement de Physique Medicale, Centre de Recherche Nucleaire d' Alger, 2 Bd Frantz Fanon BP399 Alger RP, Algiers (Algeria); Brahimi, Zakia, E-mail: zsbrahimi@yahoo.fr [Departement de Physique Medicale, Centre de Recherche Nucleaire d' Alger, 2 Bd Frantz Fanon BP399 Alger RP, Algiers (Algeria); Chami, Ahmed Chafik, E-mail: chafik_chami@yahoo.fr [Laboratoire de Sciences Nucleaires, Faculte de Physique, Universite des Sciences et de la Technologie Houari Boumedienne, BP 32 El Alia, Bab Ezzouar, Algiers (Algeria)

    2012-01-15

    Pencil beam algorithms used in computerized electron beam dose planning are usually described using the small angle multiple scattering theory. Alternatively, the pencil beams can be generated by Monte Carlo simulation of electron transport. In a previous work, the 4th version of the Electron Gamma Shower (EGS) Monte Carlo code was used to obtain dose distributions from monoenergetic electron pencil beam, with incident energy between 1 MeV and 50 MeV, interacting at the surface of a large cylindrical homogeneous water phantom. In 2000, a new version of this Monte Carlo code has been made available by the National Research Council of Canada (NRC), which includes various improvements in its electron-transport algorithms. In the present work, we were interested to see if the new physics in this version produces pencil beam dose distributions very different from those calculated with oldest one. The purpose of this study is to quantify as well as to understand these differences. We have compared a series of pencil beam dose distributions scored in cylindrical geometry, for electron energies between 1 MeV and 50 MeV calculated with two versions of the Electron Gamma Shower Monte Carlo Code. Data calculated and compared include isodose distributions, radial dose distributions and fractions of energy deposition. Our results for radial dose distributions show agreement within 10% between doses calculated by the two codes for voxels closer to the pencil beam central axis, while the differences are up to 30% for longer distances. For fractions of energy deposition, the results of the EGS4 are in good agreement (within 2%) with those calculated by EGSnrc at shallow depths for all energies, whereas a slightly worse agreement (15%) is observed at deeper distances. These differences may be mainly attributed to the different multiple scattering for electron transport adopted in these two codes and the inclusion of spin effect, which produces an increase of the effective range of

  20. SPAMCART: a code for smoothed particle Monte Carlo radiative transfer

    Science.gov (United States)

    Lomax, O.; Whitworth, A. P.

    2016-10-01

    We present a code for generating synthetic spectral energy distributions and intensity maps from smoothed particle hydrodynamics simulation snapshots. The code is based on the Lucy Monte Carlo radiative transfer method, i.e. it follows discrete luminosity packets as they propagate through a density field, and then uses their trajectories to compute the radiative equilibrium temperature of the ambient dust. The sources can be extended and/or embedded, and discrete and/or diffuse. The density is not mapped on to a grid, and therefore the calculation is performed at exactly the same resolution as the hydrodynamics. We present two example calculations using this method. First, we demonstrate that the code strictly adheres to Kirchhoff's law of radiation. Secondly, we present synthetic intensity maps and spectra of an embedded protostellar multiple system. The algorithm uses data structures that are already constructed for other purposes in modern particle codes. It is therefore relatively simple to implement.

  1. SPAMCART: a code for smoothed particle Monte Carlo radiative transfer

    CERN Document Server

    Lomax, O

    2016-01-01

    We present a code for generating synthetic SEDs and intensity maps from Smoothed Particle Hydrodynamics simulation snapshots. The code is based on the Lucy (1999) Monte Carlo Radiative Transfer method, i.e. it follows discrete luminosity packets, emitted from external and/or embedded sources, as they propagate through a density field, and then uses their trajectories to compute the radiative equilibrium temperature of the ambient dust. The density is not mapped onto a grid, and therefore the calculation is performed at exactly the same resolution as the hydrodynamics. We present two example calculations using this method. First, we demonstrate that the code strictly adheres to Kirchhoff's law of radiation. Second, we present synthetic intensity maps and spectra of an embedded protostellar multiple system. The algorithm uses data structures that are already constructed for other purposes in modern particle codes. It is therefore relatively simple to implement.

  2. ERSN-OpenMC, a Java-based GUI for OpenMC Monte Carlo code

    Directory of Open Access Journals (Sweden)

    Jaafar EL Bakkali

    2016-07-01

    Full Text Available OpenMC is a new Monte Carlo transport particle simulation code focused on solving two types of neutronic problems mainly the k-eigenvalue criticality fission source problems and external fixed fission source problems. OpenMC does not have any Graphical User Interface and the creation of one is provided by our java-based application named ERSN-OpenMC. The main feature of this application is to provide to the users an easy-to-use and flexible graphical interface to build better and faster simulations, with less effort and great reliability. Additionally, this graphical tool was developed with several features, as the ability to automate the building process of OpenMC code and related libraries as well as the users are given the freedom to customize their installation of this Monte Carlo code. A full description of the ERSN-OpenMC application is presented in this paper.

  3. Geometric Templates for Improved Tracking Performance in Monte Carlo Codes

    Science.gov (United States)

    Nease, Brian R.; Millman, David L.; Griesheimer, David P.; Gill, Daniel F.

    2014-06-01

    One of the most fundamental parts of a Monte Carlo code is its geometry kernel. This kernel not only affects particle tracking (i.e., run-time performance), but also shapes how users will input models and collect results for later analyses. A new framework based on geometric templates is proposed that optimizes performance (in terms of tracking speed and memory usage) and simplifies user input for large scale models. While some aspects of this approach currently exist in different Monte Carlo codes, the optimization aspect has not been investigated or applied. If Monte Carlo codes are to be realistically used for full core analysis and design, this type of optimization will be necessary. This paper describes the new approach and the implementation of two template types in MC21: a repeated ellipse template and a box template. Several different models are tested to highlight the performance gains that can be achieved using these templates. Though the exact gains are naturally problem dependent, results show that runtime and memory usage can be significantly reduced when using templates, even as problems reach realistic model sizes.

  4. The macro response Monte Carlo method for electron transport

    Energy Technology Data Exchange (ETDEWEB)

    Svatos, M M

    1998-09-01

    The main goal of this thesis was to prove the feasibility of basing electron depth dose calculations in a phantom on first-principles single scatter physics, in an amount of time that is equal to or better than current electron Monte Carlo methods. The Macro Response Monte Carlo (MRMC) method achieves run times that are on the order of conventional electron transport methods such as condensed history, with the potential to be much faster. This is possible because MRMC is a Local-to-Global method, meaning the problem is broken down into two separate transport calculations. The first stage is a local, in this case, single scatter calculation, which generates probability distribution functions (PDFs) to describe the electron's energy, position and trajectory after leaving the local geometry, a small sphere or "kugel" A number of local kugel calculations were run for calcium and carbon, creating a library of kugel data sets over a range of incident energies (0.25 MeV - 8 MeV) and sizes (0.025 cm to 0.1 cm in radius). The second transport stage is a global calculation, where steps that conform to the size of the kugels in the library are taken through the global geometry. For each step, the appropriate PDFs from the MRMC library are sampled to determine the electron's new energy, position and trajectory. The electron is immediately advanced to the end of the step and then chooses another kugel to sample, which continues until transport is completed. The MRMC global stepping code was benchmarked as a series of subroutines inside of the Peregrine Monte Carlo code. It was compared to Peregrine's class II condensed history electron transport package, EGS4, and MCNP for depth dose in simple phantoms having density inhomogeneities. Since the kugels completed in the library were of relatively small size, the zoning of the phantoms was scaled down from a clinical size, so that the energy deposition algorithms for spreading dose across 5-10 zones per kugel could

  5. 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.

  6. Validation of the Monte Carlo code MCNP-DSP

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.; Mihalczo, J.T. [Oak Ridge National Lab., TN (United States)

    1996-09-12

    Several calculations were performed to validate MCNP-DSP, which is a Monte Carlo code that calculates all the time and frequency analysis parameters associated with the {sup 252}Cf-source-driven time and frequency analysis method. The frequency analysis parameters are obtained in two ways: directly by Fourier transforming the detector responses and indirectly by taking the Fourier transform of the autocorrelation and cross-correlation functions. The direct and indirect Fourier processing methods were shown to produce the same frequency spectra and convergence, thus verifying the way to obtain the frequency analysis parameters from the time sequences of detector pulses. (Author).

  7. Computed radiography simulation using the Monte Carlo code MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Correa, S.C.A. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Centro Universitario Estadual da Zona Oeste (CCMAT)/UEZO, Av. Manuel Caldeira de Alvarenga, 1203, Campo Grande, 23070-200, Rio de Janeiro, RJ (Brazil); Souza, E.M. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Silva, A.X., E-mail: ademir@con.ufrj.b [PEN/COPPE-DNC/Poli CT, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Cassiano, D.H. [Instituto de Radioprotecao e Dosimetria/CNEN Av. Salvador Allende, s/n, Recreio, 22780-160, Rio de Janeiro, RJ (Brazil); Lopes, R.T. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil)

    2010-09-15

    Simulating X-ray images has been of great interest in recent years as it makes possible an analysis of how X-ray images are affected owing to relevant operating parameters. In this paper, a procedure for simulating computed radiographic images using the Monte Carlo code MCNPX is proposed. The sensitivity curve of the BaFBr image plate detector as well as the characteristic noise of a 16-bit computed radiography system were considered during the methodology's development. The results obtained confirm that the proposed procedure for simulating computed radiographic images is satisfactory, as it allows obtaining results comparable with experimental data.

  8. TRIPOLI-4{sup ®} Monte Carlo code ITER A-lite neutronic model validation

    Energy Technology Data Exchange (ETDEWEB)

    Jaboulay, Jean-Charles, E-mail: jean-charles.jaboulay@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Cayla, Pierre-Yves; Fausser, Clement [MILLENNIUM, 16 Av du Québec Silic 628, F-91945 Villebon sur Yvette (France); Damian, Frederic; Lee, Yi-Kang; Puma, Antonella Li; Trama, Jean-Christophe [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France)

    2014-10-15

    3D Monte Carlo transport codes are extensively used in neutronic analysis, especially in radiation protection and shielding analyses for fission and fusion reactors. TRIPOLI-4{sup ®} is a Monte Carlo code developed by CEA. The aim of this paper is to show its capability to model a large-scale fusion reactor with complex neutron source and geometry. A benchmark between MCNP5 and TRIPOLI-4{sup ®}, on the ITER A-lite model was carried out; neutron flux, nuclear heating in the blankets and tritium production rate in the European TBMs were evaluated and compared. The methodology to build the TRIPOLI-4{sup ®} A-lite model is based on MCAM and the MCNP A-lite model. Simplified TBMs, from KIT, were integrated in the equatorial-port. A good agreement between MCNP and TRIPOLI-4{sup ®} is shown; discrepancies are mainly included in the statistical error.

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

    CERN Document Server

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

    2001-01-01

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

  10. The Physical Models and Statistical Procedures Used in the RACER Monte Carlo Code

    Energy Technology Data Exchange (ETDEWEB)

    Sutton, T.M.; Brown, F.B.; Bischoff, F.G.; MacMillan, D.B.; Ellis, C.L.; Ward, J.T.; Ballinger, C.T.; Kelly, D.J.; Schindler, L.

    1999-07-01

    This report describes the MCV (Monte Carlo - Vectorized)Monte Carlo neutron transport code [Brown, 1982, 1983; Brown and Mendelson, 1984a]. MCV is a module in the RACER system of codes that is used for Monte Carlo reactor physics analysis. The MCV module contains all of the neutron transport and statistical analysis functions of the system, while other modules perform various input-related functions such as geometry description, material assignment, output edit specification, etc. MCV is very closely related to the 05R neutron Monte Carlo code [Irving et al., 1965] developed at Oak Ridge National Laboratory. 05R evolved into the 05RR module of the STEMB system, which was the forerunner of the RACER system. Much of the overall logic and physics treatment of 05RR has been retained and, indeed, the original verification of MCV was achieved through comparison with STEMB results. MCV has been designed to be very computationally efficient [Brown, 1981, Brown and Martin, 1984b; Brown, 1986]. It was originally programmed to make use of vector-computing architectures such as those of the CDC Cyber- 205 and Cray X-MP. MCV was the first full-scale production Monte Carlo code to effectively utilize vector-processing capabilities. Subsequently, MCV was modified to utilize both distributed-memory [Sutton and Brown, 1994] and shared memory parallelism. The code has been compiled and run on platforms ranging from 32-bit UNIX workstations to clusters of 64-bit vector-parallel supercomputers. The computational efficiency of the code allows the analyst to perform calculations using many more neutron histories than is practical with most other Monte Carlo codes, thereby yielding results with smaller statistical uncertainties. MCV also utilizes variance reduction techniques such as survival biasing, splitting, and rouletting to permit additional reduction in uncertainties. While a general-purpose neutron Monte Carlo code, MCV is optimized for reactor physics calculations. It has the

  11. Automated importance generation and biasing techniques for Monte Carlo shielding techniques by the TRIPOLI-3 code

    Energy Technology Data Exchange (ETDEWEB)

    Both, J.P.; Nimal, J.C.; Vergnaud, T. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France). Service d' Etudes des Reacteurs et de Mathematiques Appliquees)

    1990-01-01

    We discuss an automated biasing procedure for generating the parameters necessary to achieve efficient Monte Carlo biasing shielding calculations. The biasing techniques considered here are exponential transform and collision biasing deriving from the concept of the biased game based on the importance function. We use a simple model of the importance function with exponential attenuation as the distance to the detector increases. This importance function is generated on a three-dimensional mesh including geometry and with graph theory algorithms. This scheme is currently being implemented in the third version of the neutron and gamma ray transport code TRIPOLI-3. (author).

  12. OpenMC: A State-of-the-Art Monte Carlo Code for Research and Development

    Science.gov (United States)

    Romano, Paul K.; Horelik, Nicholas E.; Herman, Bryan R.; Nelson, Adam G.; Forget, Benoit; Smith, Kord

    2014-06-01

    This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently developed at the Massachusetts Institute of Technology. OpenMC uses continuous-energy cross sections and a constructive solid geometry representation, enabling high-fidelity modeling of nuclear reactors and other systems. Modern, portable input/output file formats are used in OpenMC: XML for input, and HDF5 for output. High performance parallel algorithms in OpenMC have demonstrated near-linear scaling to over 100,000 processors on modern supercomputers. Other topics discussed in this paper include plotting, CMFD acceleration, variance reduction, eigenvalue calculations, and software development processes.

  13. Experimental validation of the DPM Monte Carlo code using minimally scattered electron beams in heterogeneous media

    Science.gov (United States)

    Chetty, Indrin J.; Moran, Jean M.; Nurushev, Teamor S.; McShan, Daniel L.; Fraass, Benedick A.; Wilderman, Scott J.; Bielajew, Alex F.

    2002-06-01

    A comprehensive set of measurements and calculations has been conducted to investigate the accuracy of the Dose Planning Method (DPM) Monte Carlo code for electron beam dose calculations in heterogeneous media. Measurements were made using 10 MeV and 50 MeV minimally scattered, uncollimated electron beams from a racetrack microtron. Source distributions for the Monte Carlo calculations were reconstructed from in-air ion chamber scans and then benchmarked against measurements in a homogeneous water phantom. The in-air spatial distributions were found to have FWHM of 4.7 cm and 1.3 cm, at 100 cm from the source, for the 10 MeV and 50 MeV beams respectively. Energy spectra for the electron beams were determined by simulating the components of the microtron treatment head using the code MCNP4B. Profile measurements were made using an ion chamber in a water phantom with slabs of lung or bone-equivalent materials submerged at various depths. DPM calculations are, on average, within 2% agreement with measurement for all geometries except for the 50 MeV incident on a 6 cm lung-equivalent slab. Measurements using approximately monoenergetic, 50 MeV, 'pencil-beam'-type electrons in heterogeneous media provide conditions for maximum electronic disequilibrium and hence present a stringent test of the code's electron transport physics; the agreement noted between calculation and measurement illustrates that the DPM code is capable of accurate dose calculation even under such conditions.

  14. Reduced Fast Ion Transport Model For The Tokamak Transport Code TRANSP

    Energy Technology Data Exchange (ETDEWEB)

    Podesta,, Mario; Gorelenkova, Marina; White, Roscoe

    2014-02-28

    Fast ion transport models presently implemented in the tokamak transport code TRANSP [R. J. Hawryluk, in Physics of Plasmas Close to Thermonuclear Conditions, CEC Brussels, 1 , 19 (1980)] are not capturing important aspects of the physics associated with resonant transport caused by instabilities such as Toroidal Alfv en Eigenmodes (TAEs). This work describes the implementation of a fast ion transport model consistent with the basic mechanisms of resonant mode-particle interaction. The model is formulated in terms of a probability distribution function for the particle's steps in phase space, which is consistent with the MonteCarlo approach used in TRANSP. The proposed model is based on the analysis of fast ion response to TAE modes through the ORBIT code [R. B. White et al., Phys. Fluids 27 , 2455 (1984)], but it can be generalized to higher frequency modes (e.g. Compressional and Global Alfv en Eigenmodes) and to other numerical codes or theories.

  15. Modification of codes NUALGAM and BREMRAD. Volume 3: Statistical considerations of the Monte Carlo method

    Science.gov (United States)

    Firstenberg, H.

    1971-01-01

    The statistics are considered of the Monte Carlo method relative to the interpretation of the NUGAM2 and NUGAM3 computer code results. A numerical experiment using the NUGAM2 code is presented and the results are statistically interpreted.

  16. Reactive transport codes for subsurface environmental simulation

    NARCIS (Netherlands)

    Steefel, C.I.; Appelo, C.A.J.; Arora, B.; Kalbacher, D.; Kolditz, O.; Lagneau, V.; Lichtner, P.C.; Mayer, K.U.; Meeussen, J.C.L.; Molins, S.; Moulton, D.; Shao, D.; Simunek, J.; Spycher, N.; Yabusaki, S.B.; Yeh, G.T.

    2015-01-01

    A general description of the mathematical and numerical formulations used in modern numerical reactive transport codes relevant for subsurface environmental simulations is presented. The formulations are followed by short descriptions of commonly used and available subsurface simulators that conside

  17. A user`s manual for MASH 1.0: A Monte Carlo Adjoint Shielding Code System

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.O. [ed.

    1992-03-01

    The Monte Carlo Adjoint Shielding Code System, MASH, calculates neutron and gamma-ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air-over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system include the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. MASH is the successor to the Vehicle Code System (VCS) initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the ``dose importance`` of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response a a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user`s manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem (input data and selected output edits) for each code.

  18. A user's manual for MASH 1. 0: A Monte Carlo Adjoint Shielding Code System

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.O. (ed.)

    1992-03-01

    The Monte Carlo Adjoint Shielding Code System, MASH, calculates neutron and gamma-ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air-over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system include the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. MASH is the successor to the Vehicle Code System (VCS) initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the dose importance'' of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response a a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem (input data and selected output edits) for each code.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bordy, J.M.; Kodeli, I.; Menard, St.; Bouchet, J.L.; Renard, F.; Martin, E.; Blazy, L.; Voros, S.; Bochud, F.; Laedermann, J.P.; Beaugelin, K.; Makovicka, L.; Quiot, A.; Vermeersch, F.; Roche, H.; Perrin, M.C.; Laye, F.; Bardies, M.; Struelens, L.; Vanhavere, F.; Gschwind, R.; Fernandez, F.; Quesne, B.; Fritsch, P.; Lamart, St.; Crovisier, Ph.; Leservot, A.; Antoni, R.; Huet, Ch.; Thiam, Ch.; Donadille, L.; Monfort, M.; Diop, Ch.; Ricard, M

    2006-07-01

    The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-15

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

  1. A 3DHZETRN Code in a Spherical Uniform Sphere with Monte Carlo Verification

    Science.gov (United States)

    Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.

    2014-01-01

    The computationally efficient HZETRN code has been used in recent trade studies for lunar and Martian exploration and is currently being used in the engineering development of the next generation of space vehicles, habitats, and extra vehicular activity equipment. A new version (3DHZETRN) capable of transporting High charge (Z) and Energy (HZE) and light ions (including neutrons) under space-like boundary conditions with enhanced neutron and light ion propagation is under development. In the present report, new algorithms for light ion and neutron propagation with well-defined convergence criteria in 3D objects is developed and tested against Monte Carlo simulations to verify the solution methodology. The code will be available through the software system, OLTARIS, for shield design and validation and provides a basis for personal computer software capable of space shield analysis and optimization.

  2. Benchmarking of Heavy Ion Transport Codes

    Energy Technology Data Exchange (ETDEWEB)

    Remec, Igor [ORNL; Ronningen, Reginald M. [Michigan State University, East Lansing; Heilbronn, Lawrence [University of Tennessee, Knoxville (UTK)

    2011-01-01

    Accurate prediction of radiation fields generated by heavy ion interactions is important in medical applications, space missions, and in designing and operation of rare isotope research facilities. In recent years, several well-established computer codes in widespread use for particle and radiation transport calculations have been equipped with the capability to simulate heavy ion transport and interactions. To assess and validate these capabilities, we performed simulations of a series of benchmark-quality heavy ion experiments with the computer codes FLUKA, MARS15, MCNPX, and PHITS. We focus on the comparisons of secondary neutron production. Results are encouraging; however, further improvements in models and codes and additional benchmarking are required.

  3. Users manual for the UEDGE edge-plasma transport code

    Energy Technology Data Exchange (ETDEWEB)

    Rognlien, T D; Rensink, M E; Smith, G R

    2000-01-10

    Operational details are given for the two-dimensional UEDGE edge-plasma transport code. The model applies to nearly fully-ionized plasmas in a strong magnetic field. Equations are solved for the plasma density, velocity along the magnetic field, electron temperature, ion temperature, and electrostatic potential. In addition, fluid models of neutrals species are included or the option to couple to a Monte Carlo code description of the neutrals. Multi-species ion mixtures can be simulated. The physical equations are discretized by a finite-difference procedure, and the resulting system of algebraic equations are solved by fully-implicit techniques. The code can be used to follow time-dependent solutions or to find steady-state solutions by direct iteration.

  4. 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.

  5. A User's Manual for MASH V1.5 - A Monte Carlo Adjoint Shielding Code System

    Energy Technology Data Exchange (ETDEWEB)

    C. O. Slater; J. M. Barnes; J. O. Johnson; J.D. Drischler

    1998-10-01

    The Monte Carlo ~djoint ~ielding Code System, MASH, calculates neutron and gamma- ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air- over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system includes the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. The current version, MASH v 1.5, is the successor to the original MASH v 1.0 code system initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the "dose importance" of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response as a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem.

  6. A generic algorithm for Monte Carlo simulation of proton transport

    Energy Technology Data Exchange (ETDEWEB)

    Salvat, Francesc, E-mail: francesc.salvat@ub.edu

    2013-12-01

    A mixed (class II) algorithm for Monte Carlo simulation of the transport of protons, and other heavy charged particles, in matter is presented. The emphasis is on the electromagnetic interactions (elastic and inelastic collisions) which are simulated using strategies similar to those employed in the electron–photon code PENELOPE. Elastic collisions are described in terms of numerical differential cross sections (DCSs) in the center-of-mass frame, calculated from the eikonal approximation with the Dirac–Hartree–Fock–Slater atomic potential. The polar scattering angle is sampled by employing an adaptive numerical algorithm which allows control of interpolation errors. The energy transferred to the recoiling target atoms (nuclear stopping) is consistently described by transformation to the laboratory frame. Inelastic collisions are simulated from DCSs based on the plane–wave Born approximation (PWBA), making use of the Sternheimer–Liljequist model of the generalized oscillator strength, with parameters adjusted to reproduce (1) the electronic stopping power read from the input file, and (2) the total cross sections for impact ionization of inner subshells. The latter were calculated from the PWBA including screening and Coulomb corrections. This approach provides quite a realistic description of the energy-loss distribution in single collisions, and of the emission of X-rays induced by proton impact. The simulation algorithm can be readily modified to include nuclear reactions, when the corresponding cross sections and emission probabilities are available, and bremsstrahlung emission.

  7. A generic algorithm for Monte Carlo simulation of proton transport

    Science.gov (United States)

    Salvat, Francesc

    2013-12-01

    A mixed (class II) algorithm for Monte Carlo simulation of the transport of protons, and other heavy charged particles, in matter is presented. The emphasis is on the electromagnetic interactions (elastic and inelastic collisions) which are simulated using strategies similar to those employed in the electron-photon code PENELOPE. Elastic collisions are described in terms of numerical differential cross sections (DCSs) in the center-of-mass frame, calculated from the eikonal approximation with the Dirac-Hartree-Fock-Slater atomic potential. The polar scattering angle is sampled by employing an adaptive numerical algorithm which allows control of interpolation errors. The energy transferred to the recoiling target atoms (nuclear stopping) is consistently described by transformation to the laboratory frame. Inelastic collisions are simulated from DCSs based on the plane-wave Born approximation (PWBA), making use of the Sternheimer-Liljequist model of the generalized oscillator strength, with parameters adjusted to reproduce (1) the electronic stopping power read from the input file, and (2) the total cross sections for impact ionization of inner subshells. The latter were calculated from the PWBA including screening and Coulomb corrections. This approach provides quite a realistic description of the energy-loss distribution in single collisions, and of the emission of X-rays induced by proton impact. The simulation algorithm can be readily modified to include nuclear reactions, when the corresponding cross sections and emission probabilities are available, and bremsstrahlung emission.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  9. Experimental validation of the DPM Monte Carlo code using minimally scattered electron beams in heterogeneous media

    Energy Technology Data Exchange (ETDEWEB)

    Chetty, Indrin J. [Department of Radiation Oncology, University of Michigan, Ann Arbor, MI (United States)]. E-mail: indrin@med.umich.edu; Moran, Jean M.; Nurushev, Teamor S.; McShan, Daniel L.; Fraass, Benedick A. [Department of Radiation Oncology, University of Michigan, Ann Arbor, MI (United States); Wilderman, Scott J.; Bielajew, Alex F. [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

    2002-06-07

    A comprehensive set of measurements and calculations has been conducted to investigate the accuracy of the Dose Planning Method (DPM) Monte Carlo code for electron beam dose calculations in heterogeneous media. Measurements were made using 10 MeV and 50 MeV minimally scattered, uncollimated electron beams from a racetrack microtron. Source distributions for the Monte Carlo calculations were reconstructed from in-air ion chamber scans and then benchmarked against measurements in a homogeneous water phantom. The in-air spatial distributions were found to have FWHM of 4.7 cm and 1.3 cm, at 100 cm from the source, for the 10 MeV and 50 MeV beams respectively. Energy spectra for the electron beams were determined by simulating the components of the microtron treatment head using the code MCNP4B. Profile measurements were made using an ion chamber in a water phantom with slabs of lung or bone-equivalent materials submerged at various depths. DPM calculations are, on average, within 2% agreement with measurement for all geometries except for the 50 MeV incident on a 6 cm lung-equivalent slab. Measurements using approximately monoenergetic, 50 MeV, 'pencil-beam'-type electrons in heterogeneous media provide conditions for maximum electronic disequilibrium and hence present a stringent test of the code's electron transport physics; the agreement noted between calculation and measurement illustrates that the DPM code is capable of accurate dose calculation even under such conditions. (author)

  10. Implementation of the probability table method in a continuous-energy Monte Carlo code system

    Energy Technology Data Exchange (ETDEWEB)

    Sutton, T.M.; Brown, F.B. [Lockheed Martin Corp., Schenectady, NY (United States)

    1998-10-01

    RACER is a particle-transport Monte Carlo code that utilizes a continuous-energy treatment for neutrons and neutron cross section data. Until recently, neutron cross sections in the unresolved resonance range (URR) have been treated in RACER using smooth, dilute-average representations. This paper describes how RACER has been modified to use probability tables to treat cross sections in the URR, and the computer codes that have been developed to compute the tables from the unresolved resonance parameters contained in ENDF/B data files. A companion paper presents results of Monte Carlo calculations that demonstrate the effect of the use of probability tables versus the use of dilute-average cross sections for the URR. The next section provides a brief review of the probability table method as implemented in the RACER system. The production of the probability tables for use by RACER takes place in two steps. The first step is the generation of probability tables from the nuclear parameters contained in the ENDF/B data files. This step, and the code written to perform it, are described in Section 3. The tables produced are at energy points determined by the ENDF/B parameters and/or accuracy considerations. The tables actually used in the RACER calculations are obtained in the second step from those produced in the first. These tables are generated at energy points specific to the RACER calculation. Section 4 describes this step and the code written to implement it, as well as modifications made to RACER to enable it to use the tables. Finally, some results and conclusions are presented in Section 5.

  11. FLUKA: A Multi-Particle Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Ferrari, A.; Sala, P.R.; /CERN /INFN, Milan; Fasso, A.; /SLAC; Ranft, J.; /Siegen U.

    2005-12-14

    This report describes the 2005 version of the Fluka particle transport code. The first part introduces the basic notions, describes the modular structure of the system, and contains an installation and beginner's guide. The second part complements this initial information with details about the various components of Fluka and how to use them. It concludes with a detailed history and bibliography.

  12. Data decomposition of Monte Carlo particle transport simulations via tally servers

    Energy Technology Data Exchange (ETDEWEB)

    Romano, Paul K., E-mail: paul.k.romano@gmail.com [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 77 Massachusetts Ave., Cambridge, MA 02139 (United States); Siegel, Andrew R., E-mail: siegala@mcs.anl.gov [Argonne National Laboratory, Theory and Computing Sciences, 9700 S Cass Ave., Argonne, IL 60439 (United States); Forget, Benoit, E-mail: bforget@mit.edu [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 77 Massachusetts Ave., Cambridge, MA 02139 (United States); Smith, Kord, E-mail: kord@mit.edu [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 77 Massachusetts Ave., Cambridge, MA 02139 (United States)

    2013-11-01

    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.

  13. Domain Decomposition Strategy for Pin-wise Full-Core Monte Carlo Depletion Calculation with the Reactor Monte Carlo Code

    Directory of Open Access Journals (Sweden)

    Jingang Liang

    2016-06-01

    Full Text Available Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC codes in accomplishing pin-wise three-dimensional (3D full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.

  14. Numerical computation of discrete differential scattering cross sections for Monte Carlo charged particle transport

    Energy Technology Data Exchange (ETDEWEB)

    Walsh, Jonathan A., E-mail: walshjon@mit.edu [Department of Nuclear Science and Engineering, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-107, Cambridge, MA 02139 (United States); Palmer, Todd S. [Department of Nuclear Engineering and Radiation Health Physics, Oregon State University, 116 Radiation Center, Corvallis, OR 97331 (United States); Urbatsch, Todd J. [XTD-IDA: Theoretical Design, Integrated Design and Assessment, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2015-12-15

    Highlights: • Generation of discrete differential scattering angle and energy loss cross sections. • Gauss–Radau quadrature utilizing numerically computed cross section moments. • Development of a charged particle transport capability in the Milagro IMC code. • Integration of cross section generation and charged particle transport capabilities. - Abstract: We investigate a method for numerically generating discrete scattering cross sections for use in charged particle transport simulations. We describe the cross section generation procedure and compare it to existing methods used to obtain discrete cross sections. The numerical approach presented here is generalized to allow greater flexibility in choosing a cross section model from which to derive discrete values. Cross section data computed with this method compare favorably with discrete data generated with an existing method. Additionally, a charged particle transport capability is demonstrated in the time-dependent Implicit Monte Carlo radiative transfer code, Milagro. We verify the implementation of charged particle transport in Milagro with analytic test problems and we compare calculated electron depth–dose profiles with another particle transport code that has a validated electron transport capability. Finally, we investigate the integration of the new discrete cross section generation method with the charged particle transport capability in Milagro.

  15. Calculations for a BWR Lattice with Adjacent Gadolinium Pins Using the Monte Carlo Cell Code Serpent v.1.1.7

    Directory of Open Access Journals (Sweden)

    Diego Ferraro

    2011-01-01

    Full Text Available Monte Carlo neutron transport codes are usually used to perform criticality calculations and to solve shielding problems due to their capability to model complex systems without major approximations. However, these codes demand high computational resources. The improvement in computer capabilities leads to several new applications of Monte Carlo neutron transport codes. An interesting one is to use this method to perform cell-level fuel assembly calculations in order to obtain few group constants to be used on core calculations. In the present work the VTT recently developed Serpent v.1.1.7 cell-oriented neutronic calculation code is used to perform cell calculations of a theoretical BWR lattice benchmark with burnable poisons, and the main results are compared to reported ones and with calculations performed with Condor v.2.61, the INVAP's neutronic collision probability cell code.

  16. Data libraries as a collaborative tool across Monte Carlo codes

    CERN Document Server

    Augelli, Mauro; Han, Mincheol; Hauf, Steffen; Kim, Chan-Hyeung; Kuster, Markus; Pia, Maria Grazia; Quintieri, Lina; Saracco, Paolo; Seo, Hee; Sudhakar, Manju; Eidenspointner, Georg; Zoglauer, Andreas

    2010-01-01

    The role of data libraries in Monte Carlo simulation is discussed. A number of data libraries currently in preparation are reviewed; their data are critically examined with respect to the state-of-the-art in the respective fields. Extensive tests with respect to experimental data have been performed for the validation of their content.

  17. Overview of Particle and Heavy Ion Transport Code System PHITS

    Science.gov (United States)

    Sato, Tatsuhiko; Niita, Koji; Matsuda, Norihiro; Hashimoto, Shintaro; Iwamoto, Yosuke; Furuta, Takuya; Noda, Shusaku; Ogawa, Tatsuhiko; Iwase, Hiroshi; Nakashima, Hiroshi; Fukahori, Tokio; Okumura, Keisuke; Kai, Tetsuya; Chiba, Satoshi; Sihver, Lembit

    2014-06-01

    A general purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS, is being developed through the collaboration of several institutes in Japan and Europe. The Japan Atomic Energy Agency is responsible for managing the entire project. PHITS can deal with the transport of nearly all particles, including neutrons, protons, heavy ions, photons, and electrons, over wide energy ranges using various nuclear reaction models and data libraries. It is written in Fortran language and can be executed on almost all computers. All components of PHITS such as its source, executable and data-library files are assembled in one package and then distributed to many countries via the Research organization for Information Science and Technology, the Data Bank of the Organization for Economic Co-operation and Development's Nuclear Energy Agency, and the Radiation Safety Information Computational Center. More than 1,000 researchers have been registered as PHITS users, and they apply the code to various research and development fields such as nuclear technology, accelerator design, medical physics, and cosmic-ray research. This paper briefly summarizes the physics models implemented in PHITS, and introduces some important functions useful for specific applications, such as an event generator mode and beam transport functions.

  18. Applicability of the SCALE code system to MOX fuel transport systems for criticality safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Toshihiro; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hayashi, Toshiaki; Takasugi, Masahiro; Natsume, Toshihiro; Tsuda, Kazuaki

    1996-11-01

    In order to ascertain feasibilities of the SCALE code system for MOX fuel transport systems, criticality analyses were performed for MOX fuel (Pu enrichment; 3.0 wt.%) criticality experiments at JAERI`s TCA and for infinite fuel rod arrays as parameters of Pu enrichment and lattice pitch. The comparison with a combination of the continuous energy Monte Carlo code MCNP and JENDL-3.2 indicated that the SCALE code system with GAM-THERMOS 123-group library can produce feasible results. Though HANSEN-ROACH 16-group library gives poorer results for MOS fuel transport systems, the errors are conservative except for high enriched fuels. (author)

  19. Monte Carlo simulations of charge transport in heterogeneous organic semiconductors

    Science.gov (United States)

    Aung, Pyie Phyo; Khanal, Kiran; Luettmer-Strathmann, Jutta

    2015-03-01

    The efficiency of organic solar cells depends on the morphology and electronic properties of the active layer. Research teams have been experimenting with different conducting materials to achieve more efficient solar panels. In this work, we perform Monte Carlo simulations to study charge transport in heterogeneous materials. We have developed a coarse-grained lattice model of polymeric photovoltaics and use it to generate active layers with ordered and disordered regions. We determine carrier mobilities for a range of conditions to investigate the effect of the morphology on charge transport.

  20. Progress and status of the OpenMC Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Romano, P. K.; Herman, B. R.; Horelik, N. E.; Forget, B.; Smith, K. [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States); Siegel, A. R. [Argonne National Laboratory, Theory and Computing Sciences and Nuclear Engineering Division (United States)

    2013-07-01

    The present work describes the latest advances and progress in the development of the OpenMC Monte Carlo code, an open-source code originating from the Massachusetts Institute of Technology. First, an overview of the development workflow of OpenMC is given. Various enhancements to the code such as real-time XML input validation, state points, plotting, OpenMP threading, and coarse mesh finite difference acceleration are described. (authors)

  1. Monte Carlo Neutrino Transport Through Remnant Disks from Neutron Star Mergers

    CERN Document Server

    Richers, S; O'Connor, Evan; Fernandez, Rodrigo; Ott, Christian

    2015-01-01

    We present Sedonu, a new open source, steady-state, special relativistic Monte Carlo (MC) neutrino transport code, available at bitbucket.org/srichers/sedonu. The code calculates the energy- and angle-dependent neutrino distribution function on fluid backgrounds of any number of spatial dimensions, calculates the rates of change of fluid internal energy and electron fraction, and solves for the equilibrium fluid temperature and electron fraction. We apply this method to snapshots from two dimensional simulations of accretion disks left behind by binary neutron star mergers, varying the input physics and comparing to the results obtained with a leakage scheme for the case of a central black hole and a central hypermassive neutron star. Neutrinos are guided away from the densest regions of the disk and escape preferentially around 45 degrees from the equatorial plane. Neutrino heating is strengthened by MC transport a few scale heights above the disk midplane near the innermost stable circular orbit, potentiall...

  2. Monte Carlo modelling of positron transport in real world applications

    Science.gov (United States)

    Marjanović, S.; Banković, A.; Šuvakov, M.; Petrović, Z. Lj

    2014-05-01

    Due to the unstable nature of positrons and their short lifetime, it is difficult to obtain high positron particle densities. This is why the Monte Carlo simulation technique, as a swarm method, is very suitable for modelling most of the current positron applications involving gaseous and liquid media. The ongoing work on the measurements of cross-sections for positron interactions with atoms and molecules and swarm calculations for positrons in gasses led to the establishment of good cross-section sets for positron interaction with gasses commonly used in real-world applications. Using the standard Monte Carlo technique and codes that can follow both low- (down to thermal energy) and high- (up to keV) energy particles, we are able to model different systems directly applicable to existing experimental setups and techniques. This paper reviews the results on modelling Surko-type positron buffer gas traps, application of the rotating wall technique and simulation of positron tracks in water vapor as a substitute for human tissue, and pinpoints the challenges in and advantages of applying Monte Carlo simulations to these systems.

  3. Benchmarking of Proton Transport in Super Monte Carlo Simulation Program

    Science.gov (United States)

    Wang, Yongfeng; Li, Gui; Song, Jing; Zheng, Huaqing; Sun, Guangyao; Hao, Lijuan; Wu, Yican

    2014-06-01

    The Monte Carlo (MC) method has been traditionally applied in nuclear design and analysis due to its capability of dealing with complicated geometries and multi-dimensional physics problems as well as obtaining accurate results. The Super Monte Carlo Simulation Program (SuperMC) is developed by FDS Team in China for fusion, fission, and other nuclear applications. The simulations of radiation transport, isotope burn-up, material activation, radiation dose, and biology damage could be performed using SuperMC. Complicated geometries and the whole physical process of various types of particles in broad energy scale can be well handled. Bi-directional automatic conversion between general CAD models and full-formed input files of SuperMC is supported by MCAM, which is a CAD/image-based automatic modeling program for neutronics and radiation transport simulation. Mixed visualization of dynamical 3D dataset and geometry model is supported by RVIS, which is a nuclear radiation virtual simulation and assessment system. Continuous-energy cross section data from hybrid evaluated nuclear data library HENDL are utilized to support simulation. Neutronic fixed source and critical design parameters calculates for reactors of complex geometry and material distribution based on the transport of neutron and photon have been achieved in our former version of SuperMC. Recently, the proton transport has also been intergrated in SuperMC in the energy region up to 10 GeV. The physical processes considered for proton transport include electromagnetic processes and hadronic processes. The electromagnetic processes include ionization, multiple scattering, bremsstrahlung, and pair production processes. Public evaluated data from HENDL are used in some electromagnetic processes. In hadronic physics, the Bertini intra-nuclear cascade model with exitons, preequilibrium model, nucleus explosion model, fission model, and evaporation model are incorporated to treat the intermediate energy nuclear

  4. Longitudinal development of extensive air showers: hybrid code SENECA and full Monte Carlo

    CERN Document Server

    Ortiz, J A; De Souza, V; Ortiz, Jeferson A.; Tanco, Gustavo Medina

    2004-01-01

    New experiments, exploring the ultra-high energy tail of the cosmic ray spectrum with unprecedented detail, are exerting a severe pressure on extensive air hower modeling. Detailed fast codes are in need in order to extract and understand the richness of information now available. Some hybrid simulation codes have been proposed recently to this effect (e.g., the combination of the traditional Monte Carlo scheme and system of cascade equations or pre-simulated air showers). In this context, we explore the potential of SENECA, an efficient hybrid tridimensional simulation code, as a valid practical alternative to full Monte Carlo simulations of extensive air showers generated by ultra-high energy cosmic rays. We extensively compare hybrid method with the traditional, but time consuming, full Monte Carlo code CORSIKA which is the de facto standard in the field. The hybrid scheme of the SENECA code is based on the simulation of each particle with the traditional Monte Carlo method at two steps of the shower devel...

  5. Longitudinal development of extensive air showers: Hybrid code SENECA and full Monte Carlo

    Science.gov (United States)

    Ortiz, Jeferson A.; Medina-Tanco, Gustavo; de Souza, Vitor

    2005-06-01

    New experiments, exploring the ultra-high energy tail of the cosmic ray spectrum with unprecedented detail, are exerting a severe pressure on extensive air shower modelling. Detailed fast codes are in need in order to extract and understand the richness of information now available. Some hybrid simulation codes have been proposed recently to this effect (e.g., the combination of the traditional Monte Carlo scheme and system of cascade equations or pre-simulated air showers). In this context, we explore the potential of SENECA, an efficient hybrid tri-dimensional simulation code, as a valid practical alternative to full Monte Carlo simulations of extensive air showers generated by ultra-high energy cosmic rays. We extensively compare hybrid method with the traditional, but time consuming, full Monte Carlo code CORSIKA which is the de facto standard in the field. The hybrid scheme of the SENECA code is based on the simulation of each particle with the traditional Monte Carlo method at two steps of the shower development: the first step predicts the large fluctuations in the very first particle interactions at high energies while the second step provides a well detailed lateral distribution simulation of the final stages of the air shower. Both Monte Carlo simulation steps are connected by a cascade equation system which reproduces correctly the hadronic and electromagnetic longitudinal profile. We study the influence of this approach on the main longitudinal characteristics of proton, iron nucleus and gamma induced air showers and compare the predictions of the well known CORSIKA code using the QGSJET hadronic interaction model.

  6. Calculation of electron and isotopes dose point kernels with FLUKA Monte Carlo code for dosimetry in nuclear medicine therapy

    CERN Document Server

    Mairani, A; Valente, M; Battistoni, G; Botta, F; Pedroli, G; Ferrari, A; Cremonesi, M; Di Dia, A; Ferrari, M; Fasso, A

    2011-01-01

    Purpose: The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, FLUKA Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, FLUKA has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one. Methods: FLUKA DPKS have been calculated in both water and compact bone for monoenergetic electrons (10-3 MeV) and for beta emitting isotopes commonly used for therapy ((89)Sr, (90)Y, (131)I, (153)Sm, (177)Lu, (186)Re, and (188)Re). Point isotropic...

  7. Modeling Monte Carlo of multileaf collimators using the code GEANT4

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Alex C.H.; Lima, Fernando R.A., E-mail: oliveira.ach@yahoo.com, E-mail: falima@cnen.gov.br [Centro Regional de Ciencias Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil); Lima, Luciano S.; Vieira, Jose W., E-mail: lusoulima@yahoo.com.br [Instituto Federal de Educacao, Ciencia e Tecnologia de Pernambuco (IFPE), Recife, PE (Brazil)

    2014-07-01

    Radiotherapy uses various techniques and equipment for local treatment of cancer. The equipment most often used in radiotherapy to the patient irradiation is linear accelerator (Linac). Among the many algorithms developed for evaluation of dose distributions in radiotherapy planning, the algorithms based on Monte Carlo (MC) methods have proven to be very promising in terms of accuracy by providing more realistic results. The MC simulations for applications in radiotherapy are divided into two parts. In the first, the simulation of the production of the radiation beam by the Linac is performed and then the phase space is generated. The phase space contains information such as energy, position, direction, etc. of millions of particles (photons, electrons, positrons). In the second part the simulation of the transport of particles (sampled phase space) in certain configurations of irradiation field is performed to assess the dose distribution in the patient (or phantom). Accurate modeling of the Linac head is of particular interest in the calculation of dose distributions for intensity modulated radiation therapy (IMRT), where complex intensity distributions are delivered using a multileaf collimator (MLC). The objective of this work is to describe a methodology for modeling MC of MLCs using code Geant4. To exemplify this methodology, the Varian Millennium 120-leaf MLC was modeled, whose physical description is available in BEAMnrc Users Manual (20 11). The dosimetric characteristics (i.e., penumbra, leakage, and tongue-and-groove effect) of this MLC were evaluated. The results agreed with data published in the literature concerning the same MLC. (author)

  8. Simulation of the Mg(Ar) ionization chamber currents by different Monte Carlo codes in benchmark gamma fields

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Yi-Chun [Department of Biomedical Engineering and Environmental Sciences, National Tsing Hua University, Taiwan (China); Liu, Yuan-Hao, E-mail: yhl.taiwan@gmail.com [Boron Neutron Capture Therapy Center, Nuclear Science and Technology Development Center, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Road, Hsinchu City 30013, Taiwan (China); Nievaart, Sander [Institute for Energy, Joint Research Centre, European Commission, Petten (Netherlands); Chen, Yen-Fu [Department of Engineering and System Science, National Tsing Hua University, Taiwan (China); Wu, Shu-Wei; Chou, Wen-Tsae [Department of Biomedical Engineering and Environmental Sciences, National Tsing Hua University, Taiwan (China); Jiang, Shiang-Huei [Institute of Nuclear Engineering and Science, National Tsing Hua University, Taiwan (China)

    2011-10-01

    High energy photon (over 10 MeV) and neutron beams adopted in radiobiology and radiotherapy always produce mixed neutron/gamma-ray fields. The Mg(Ar) ionization chambers are commonly applied to determine the gamma-ray dose because of its neutron insensitive characteristic. Nowadays, many perturbation corrections for accurate dose estimation and lots of treatment planning systems are based on Monte Carlo technique. The Monte Carlo codes EGSnrc, FLUKA, GEANT4, MCNP5, and MCNPX were used to evaluate energy dependent response functions of the Exradin M2 Mg(Ar) ionization chamber to a parallel photon beam with mono-energies from 20 keV to 20 MeV. For the sake of validation, measurements were carefully performed in well-defined (a) primary M-100 X-ray calibration field, (b) primary {sup 60}Co calibration beam, (c) 6-MV, and (d) 10-MV therapeutic beams in hospital. At energy region below 100 keV, MCNP5 and MCNPX both had lower responses than other codes. For energies above 1 MeV, the MCNP ITS-mode greatly resembled other three codes and the differences were within 5%. Comparing to the measured currents, MCNP5 and MCNPX using ITS-mode had perfect agreement with the {sup 60}Co, and 10-MV beams. But at X-ray energy region, the derivations reached 17%. This work shows us a better insight into the performance of different Monte Carlo codes in photon-electron transport calculation. Regarding the application of the mixed field dosimetry like BNCT, MCNP with ITS-mode is recognized as the most suitable tool by this work.

  9. NEPHTIS: 2D/3D validation elements using MCNP4c and TRIPOLI4 Monte-Carlo codes

    Energy Technology Data Exchange (ETDEWEB)

    Courau, T.; Girardi, E. [EDF R and D/SINETICS, 1av du General de Gaulle, F92141 Clamart CEDEX (France); Damian, F.; Moiron-Groizard, M. [DEN/DM2S/SERMA/LCA, CEA Saclay, F91191 Gif-sur-Yvette CEDEX (France)

    2006-07-01

    High Temperature Reactors (HTRs) appear as a promising concept for the next generation of nuclear power applications. The CEA, in collaboration with AREVA-NP and EDF, is developing a core modeling tool dedicated to the prismatic block-type reactor. NEPHTIS (Neutronics Process for HTR Innovating System) is a deterministic codes system based on a standard two-steps Transport-Diffusion approach (APOLLO2/CRONOS2). Validation of such deterministic schemes usually relies on Monte-Carlo (MC) codes used as a reference. However, when dealing with large HTR cores the fission source stabilization is rather poor with MC codes. In spite of this, it is shown in this paper that MC simulations may be used as a reference for a wide range of configurations. The first part of the paper is devoted to 2D and 3D MC calculations of a HTR core with control devices. Comparisons between MCNP4c and TRIPOLI4 MC codes are performed and show very consistent results. Finally, the last part of the paper is devoted to the code to code validation of the NEPHTIS deterministic scheme. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Bekar, Kursat B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, Thomas Martin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Weber, Charles F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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.

  11. A new Monte Carlo code for absorption simulation of laser-skin tissue interaction

    Institute of Scientific and Technical Information of China (English)

    Afshan Shirkavand; Saeed Sarkar; Marjaneh Hejazi; Leila Ataie-Fashtami; Mohammad Reza Alinaghizadeh

    2007-01-01

    In laser clinical applications, the process of photon absorption and thermal energy diffusion in the target tissue and its surrounding tissue during laser irradiation are crucial. Such information allows the selection of proper operating parameters such as laser power, and exposure time for optimal therapeutic. The Monte Carlo method is a useful tool for studying laser-tissue interaction and simulation of energy absorption in tissue during laser irradiation. We use the principles of this technique and write a new code with MATLAB 6.5, and then validate it against Monte Carlo multi layer (MCML) code. The new code is proved to be with good accuracy. It can be used to calculate the total power bsorbed in the region of interest. This can be combined for heat modelling with other computerized programs.

  12. Applications of FLUKA Monte Carlo code for nuclear and accelerator physics

    CERN Document Server

    Battistoni, Giuseppe; Brugger, Markus; Campanella, Mauro; Carboni, Massimo; Empl, Anton; Fasso, Alberto; Gadioli, Ettore; Cerutti, Francesco; Ferrari, Alfredo; Ferrari, Anna; Lantz, Matthias; Mairani, Andrea; Margiotta, M; Morone, Christina; Muraro, Silvia; Parodi, Katerina; Patera, Vincenzo; Pelliccioni, Maurizio; Pinsky, Lawrence; Ranft, Johannes; Roesler, Stefan; Rollet, Sofia; Sala, Paola R; Santana, Mario; Sarchiapone, Lucia; Sioli, Maximiliano; Smirnov, George; Sommerer, Florian; Theis, Christian; Trovati, Stefania; Villari, R; Vincke, Heinz; Vincke, Helmut; Vlachoudis, Vasilis; Vollaire, Joachim; Zapp, Neil

    2011-01-01

    FLUKA is a general purpose Monte Carlo code capable of handling all radiation components from thermal energies (for neutrons) or 1keV (for all other particles) to cosmic ray energies and can be applied in many different fields. Presently the code is maintained on Linux. The validity of the physical models implemented in FLUKA has been benchmarked against a variety of experimental data over a wide energy range, from accelerator data to cosmic ray showers in the Earth atmosphere. FLUKA is widely used for studies related both to basic research and to applications in particle accelerators, radiation protection and dosimetry, including the specific issue of radiation damage in space missions, radiobiology (including radiotherapy) and cosmic ray calculations. After a short description of the main features that make FLUKA valuable for these topics, the present paper summarizes some of the recent applications of the FLUKA Monte Carlo code in the nuclear as well high energy physics. In particular it addresses such top...

  13. Accuracy and convergence of coupled finite-volume/Monte Carlo codes for plasma edge simulations of nuclear fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ghoos, K., E-mail: kristel.ghoos@kuleuven.be [KU Leuven, Department of Mechanical Engineering, Celestijnenlaan 300A, 3001 Leuven (Belgium); Dekeyser, W. [KU Leuven, Department of Mechanical Engineering, Celestijnenlaan 300A, 3001 Leuven (Belgium); Samaey, G. [KU Leuven, Department of Computer Science, Celestijnenlaan 200A, 3001 Leuven (Belgium); Börner, P. [Institute of Energy and Climate Research (IEK-4), FZ Jülich GmbH, D-52425 Jülich (Germany); Baelmans, M. [KU Leuven, Department of Mechanical Engineering, Celestijnenlaan 300A, 3001 Leuven (Belgium)

    2016-10-01

    The plasma and neutral transport in the plasma edge of a nuclear fusion reactor is usually simulated using coupled finite volume (FV)/Monte Carlo (MC) codes. However, under conditions of future reactors like ITER and DEMO, convergence issues become apparent. This paper examines the convergence behaviour and the numerical error contributions with a simplified FV/MC model for three coupling techniques: Correlated Sampling, Random Noise and Robbins Monro. Also, practical procedures to estimate the errors in complex codes are proposed. Moreover, first results with more complex models show that an order of magnitude speedup can be achieved without any loss in accuracy by making use of averaging in the Random Noise coupling technique.

  14. SYMTRAN - A Time-dependent Symmetric Tandem Mirror Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Hua, D; Fowler, T

    2004-06-15

    A time-dependent version of the steady-state radial transport model in symmetric tandem mirrors in Ref. [1] has been coded up and first tests performed. Our code, named SYMTRAN, is an adaptation of the earlier SPHERE code for spheromaks, now modified for tandem mirror physics. Motivated by Post's new concept of kinetic stabilization of symmetric mirrors, it is an extension of the earlier TAMRAC rate-equation code omitting radial transport [2], which successfully accounted for experimental results in TMX. The SYMTRAN code differs from the earlier tandem mirror radial transport code TMT in that our code is focused on axisymmetric tandem mirrors and classical diffusion, whereas TMT emphasized non-ambipolar transport in TMX and MFTF-B due to yin-yang plugs and non-symmetric transitions between the plugs and axisymmetric center cell. Both codes exhibit interesting but different non-linear behavior.

  15. SU-E-T-254: Optimization of GATE and PHITS Monte Carlo Code Parameters for Uniform Scanning Proton Beam Based On Simulation with FLUKA General-Purpose Code

    Energy Technology Data Exchange (ETDEWEB)

    Kurosu, K [Department of Radiation Oncology, Osaka University Graduate School of Medicine, Osaka (Japan); Department of Medical Physics ' Engineering, Osaka University Graduate School of Medicine, Osaka (Japan); Takashina, M; Koizumi, M [Department of Medical Physics ' Engineering, Osaka University Graduate School of Medicine, Osaka (Japan); Das, I; Moskvin, V [Department of Radiation Oncology, Indiana University School of Medicine, Indianapolis, IN (United States)

    2014-06-01

    Purpose: Monte Carlo codes are becoming important tools for proton beam dosimetry. However, the relationships between the customizing parameters and percentage depth dose (PDD) of GATE and PHITS codes have not been reported which are studied for PDD and proton range compared to the FLUKA code and the experimental data. Methods: The beam delivery system of the Indiana University Health Proton Therapy Center was modeled for the uniform scanning beam in FLUKA and transferred identically into GATE and PHITS. This computational model was built from the blue print and validated with the commissioning data. Three parameters evaluated are the maximum step size, cut off energy and physical and transport model. The dependence of the PDDs on the customizing parameters was compared with the published results of previous studies. Results: The optimal parameters for the simulation of the whole beam delivery system were defined by referring to the calculation results obtained with each parameter. Although the PDDs from FLUKA and the experimental data show a good agreement, those of GATE and PHITS obtained with our optimal parameters show a minor discrepancy. The measured proton range R90 was 269.37 mm, compared to the calculated range of 269.63 mm, 268.96 mm, and 270.85 mm with FLUKA, GATE and PHITS, respectively. Conclusion: We evaluated the dependence of the results for PDDs obtained with GATE and PHITS Monte Carlo generalpurpose codes on the customizing parameters by using the whole computational model of the treatment nozzle. The optimal parameters for the simulation were then defined by referring to the calculation results. The physical model, particle transport mechanics and the different geometrybased descriptions need accurate customization in three simulation codes to agree with experimental data for artifact-free Monte Carlo simulation. This study was supported by Grants-in Aid for Cancer Research (H22-3rd Term Cancer Control-General-043) from the Ministry of Health

  16. Description of Transport Codes for Space Radiation Shielding

    Science.gov (United States)

    Kim, Myung-Hee Y.; Wilson, John W.; Cucinotta, Francis A.

    2011-01-01

    This slide presentation describes transport codes and their use for studying and designing space radiation shielding. When combined with risk projection models radiation transport codes serve as the main tool for study radiation and designing shielding. There are three criteria for assessing the accuracy of transport codes: (1) Ground-based studies with defined beams and material layouts, (2) Inter-comparison of transport code results for matched boundary conditions and (3) Comparisons to flight measurements. These three criteria have a very high degree with NASA's HZETRN/QMSFRG.

  17. DgSMC-B code: A robust and autonomous direct simulation Monte Carlo code for arbitrary geometries

    Science.gov (United States)

    Kargaran, H.; Minuchehr, A.; Zolfaghari, A.

    2016-07-01

    In this paper, we describe the structure of a new Direct Simulation Monte Carlo (DSMC) code that takes advantage of combinatorial geometry (CG) to simulate any rarefied gas flows Medias. The developed code, called DgSMC-B, has been written in FORTRAN90 language with capability of parallel processing using OpenMP framework. The DgSMC-B is capable of handling 3-dimensional (3D) geometries, which is created with first-and second-order surfaces. It performs independent particle tracking for the complex geometry without the intervention of mesh. In addition, it resolves the computational domain boundary and volume computing in border grids using hexahedral mesh. The developed code is robust and self-governing code, which does not use any separate code such as mesh generators. The results of six test cases have been presented to indicate its ability to deal with wide range of benchmark problems with sophisticated geometries such as airfoil NACA 0012. The DgSMC-B code demonstrates its performance and accuracy in a variety of problems. The results are found to be in good agreement with references and experimental data.

  18. Current status of high energy nucleon-meson transport code

    Energy Technology Data Exchange (ETDEWEB)

    Takada, Hiroshi; Sasa, Toshinobu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Current status of design code of accelerator (NMTC/JAERI code), outline of physical model and evaluation of accuracy of code were reported. To evaluate the nuclear performance of accelerator and strong spallation neutron origin, the nuclear reaction between high energy proton and target nuclide and behaviors of various produced particles are necessary. The nuclear design of spallation neutron system used a calculation code system connected the high energy nucleon{center_dot}meson transport code and the neutron{center_dot}photon transport code. NMTC/JAERI is described by the particle evaporation process under consideration of competition reaction of intranuclear cascade and fission process. Particle transport calculation was carried out for proton, neutron, {pi}- and {mu}-meson. To verify and improve accuracy of high energy nucleon-meson transport code, data of spallation and spallation neutron fragment by the integral experiment were collected. (S.Y.)

  19. Evaluation of CASMO-3 and HELIOS for Fuel Assembly Analysis from Monte Carlo Code

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Hyung Jin; Song, Jae Seung; Lee, Chung Chan

    2007-05-15

    This report presents a study comparing deterministic lattice physics calculations with Monte Carlo calculations for LWR fuel pin and assembly problems. The study has focused on comparing results from the lattice physics code CASMO-3 and HELIOS against those from the continuous-energy Monte Carlo code McCARD. The comparisons include k{sub inf}, isotopic number densities, and pin power distributions. The CASMO-3 and HELIOS calculations for the k{sub inf}'s of the LWR fuel pin problems show good agreement with McCARD within 956pcm and 658pcm, respectively. For the assembly problems with Gadolinia burnable poison rods, the largest difference between the k{sub inf}'s is 1463pcm with CASMO-3 and 1141pcm with HELIOS. RMS errors for the pin power distributions of CASMO-3 and HELIOS are within 1.3% and 1.5%, respectively.

  20. The Serpent Monte Carlo Code: Status, Development and Applications in 2013

    Science.gov (United States)

    Leppänen, Jaakko; Pusa, Maria; Viitanen, Tuomas; Valtavirta, Ville; Kaltiaisenaho, Toni

    2014-06-01

    The Serpent Monte Carlo reactor physics burnup calculation code has been developed at VTT Technical Research Centre of Finland since 2004, and is currently used in 100 universities and research organizations around the world. This paper presents the brief history of the project, together with the currently available methods and capabilities and plans for future work. Typical user applications are introduced in the form of a summary review on Serpent-related publications over the past few years.

  1. ASCOT: redesigned Monte Carlo code for simulations of minority species in tokamak plasmas

    CERN Document Server

    Hirvijoki, Eero; Koskela, Tuomas; Kurki-Suonio, Taina; Miettunen, Juho; Sipilä, Seppo; Snicker, Antti; Äkäslompolo, Simppa

    2013-01-01

    A comprehensive description of methods for Monte Carlo studies of fast ions and impurity species in tokamak plasmas is presented. The described methods include Hamiltonian orbit-following in particle and guiding center phase space, test particle or guiding center solution of the kinetic equation applying stochastic differential equations in the presence of Coulomb collisions, Neoclassical tearing modes and Alfv\\'en eigenmodes as electromagnetic perturbations relevant for fast ions, together with plasma flow and atomic reactions relevant for impurity studies. Applying the methods, a complete reimplementation of a well-established minority species code is carried out as a response both to the increase in computing power during the last twenty years and to the weakly structured growth of the previous code which has made implementation of additional models impractical. Also, a thorough benchmark between the previous code and the reimplementation is accomplished, showing good agreement between the codes.

  2. Validation of deterministic and Monte Carlo codes for neutronics calculation of the IRT-type research reactor

    Science.gov (United States)

    Shchurovskaya, M. V.; Alferov, V. P.; Geraskin, N. I.; Radaev, A. I.

    2017-01-01

    The results of the validation of a research reactor calculation using Monte Carlo and deterministic codes against experimental data and based on code-to-code comparison are presented. The continuous energy Monte Carlo code MCU-PTR and the nodal diffusion-based deterministic code TIGRIS were used for full 3-D calculation of the IRT MEPhI research reactor. The validation included the investigations for the reactor with existing high enriched uranium (HEU, 90 w/o) fuel and low enriched uranium (LEU, 19.7 w/o, U-9%Mo) fuel.

  3. 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)

  4. Radiation Transport for Explosive Outflows: A Multigroup Hybrid Monte Carlo Method

    Science.gov (United States)

    Wollaeger, Ryan T.; van Rossum, Daniel R.; Graziani, Carlo; Couch, Sean M.; Jordan, George C., IV; Lamb, Donald Q.; Moses, Gregory A.

    2013-12-01

    We explore Implicit Monte Carlo (IMC) and discrete diffusion Monte Carlo (DDMC) for radiation transport in high-velocity outflows with structured opacity. The IMC method is a stochastic computational technique for nonlinear radiation transport. IMC is partially implicit in time and may suffer in efficiency when tracking MC particles through optically thick materials. DDMC accelerates IMC in diffusive domains. Abdikamalov extended IMC and DDMC to multigroup, velocity-dependent transport with the intent of modeling neutrino dynamics in core-collapse supernovae. Densmore has also formulated a multifrequency extension to the originally gray DDMC method. We rigorously formulate IMC and DDMC over a high-velocity Lagrangian grid for possible application to photon transport in the post-explosion phase of Type Ia supernovae. This formulation includes an analysis that yields an additional factor in the standard IMC-to-DDMC spatial interface condition. To our knowledge the new boundary condition is distinct from others presented in prior DDMC literature. The method is suitable for a variety of opacity distributions and may be applied to semi-relativistic radiation transport in simple fluids and geometries. Additionally, we test the code, called SuperNu, using an analytic solution having static material, as well as with a manufactured solution for moving material with structured opacities. Finally, we demonstrate with a simple source and 10 group logarithmic wavelength grid that IMC-DDMC performs better than pure IMC in terms of accuracy and speed when there are large disparities between the magnitudes of opacities in adjacent groups. We also present and test our implementation of the new boundary condition.

  5. MCNPX Monte Carlo simulations of particle transport in SiC semiconductor detectors of fast neutrons

    Science.gov (United States)

    Sedlačková, K.; Zat'ko, B.; Šagátová, A.; Pavlovič, M.; Nečas, V.; Stacho, M.

    2014-05-01

    The aim of this paper was to investigate particle transport properties of a fast neutron detector based on silicon carbide. MCNPX (Monte Carlo N-Particle eXtended) code was used in our study because it allows seamless particle transport, thus not only interacting neutrons can be inspected but also secondary particles can be banked for subsequent transport. Modelling of the fast-neutron response of a SiC detector was carried out for fast neutrons produced by 239Pu-Be source with the mean energy of about 4.3 MeV. Using the MCNPX code, the following quantities have been calculated: secondary particle flux densities, reaction rates of elastic/inelastic scattering and other nuclear reactions, distribution of residual ions, deposited energy and energy distribution of pulses. The values of reaction rates calculated for different types of reactions and resulting energy deposition values showed that the incident neutrons transfer part of the carried energy predominantly via elastic scattering on silicon and carbon atoms. Other fast-neutron induced reactions include inelastic scattering and nuclear reactions followed by production of α-particles and protons. Silicon and carbon recoil atoms, α-particles and protons are charged particles which contribute to the detector response. It was demonstrated that although the bare SiC material can register fast neutrons directly, its detection efficiency can be enlarged if it is covered by an appropriate conversion layer. Comparison of the simulation results with experimental data was successfully accomplished.

  6. Introduction to the simulation with MCNP Monte Carlo code and its applications in Medical Physics; Introduccion a la simulacion con el codigo de Monte Carlo MCNP y sus aplicaciones en Fisica Medica

    Energy Technology Data Exchange (ETDEWEB)

    Parreno Z, F.; Paucar J, R.; Picon C, C. [Instituto Peruano de Energia Nuclear, Av. Canada 1470, San Borja, Lima 41 (Peru)

    1998-12-31

    The simulation by Monte Carlo is tool which Medical Physics counts with it for the development of its research, the interest by this tool is growing, as we may observe in the main scientific journals for the years 1995-1997 where more than 27 % of the papers treat over Monte Carlo and/or its applications in the radiation transport.In the Peruvian Institute of Nuclear Energy we are implementing and making use of the MCNP4 and EGS4 codes. In this work are presented the general features of the Monte Carlo method and its more useful applications in Medical Physics. Likewise, it is made a simulation of the calculation of isodose curves in an interstitial treatment with Ir-192 wires in a mammary gland carcinoma. (Author)

  7. GPU-Accelerated Monte Carlo Electron Transport Methods: Development and Application for Radiation Dose Calculations Using Six GPU cards

    Science.gov (United States)

    Su, Lin; Du, Xining; Liu, Tianyu; Xu, X. George

    2014-06-01

    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 testbed for emerging heterogeneous high performance computers that utilize accelerators such as GPUs. 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. For photon part, photoelectric effect, Compton scattering and pair production were modeled. Voxelized geometry was supported. A serial CPU 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, 6x106 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.

  8. Monte-Carlo code calculation of 3D reactor core model with usage of burnt fuel isotopic compositions, obtained by engineering codes

    Energy Technology Data Exchange (ETDEWEB)

    Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

    2016-09-15

    A burn-up calculation of large systems by Monte-Carlo code (MCU) is complex process and it requires large computational costs. Previously prepared isotopic compositions are proposed to be used for the Monte-Carlo code calculations of different system states with burnt fuel. Isotopic compositions are calculated by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by the engineering codes (TVS-M, BIPR-7A and PERMAK-A). The multiplication factors and power distributions of FAs from a 3-D reactor core are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The separate conditions of the burnt core are observed. The results of MCU calculations were compared with those that were obtained by engineering codes.

  9. Energy Conservation Tests of a Coupled Kinetic-kinetic Plasma-neutral Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Stotler, D. P.; Chang, C. S.; Ku, S. H.; Lang, J.; Park, G.

    2012-08-29

    A Monte Carlo neutral transport routine, based on DEGAS2, has been coupled to the guiding center ion-electron-neutral neoclassical PIC code XGC0 to provide a realistic treatment of neutral atoms and molecules in the tokamak edge plasma. The DEGAS2 routine allows detailed atomic physics and plasma-material interaction processes to be incorporated into these simulations. The spatial pro le of the neutral particle source used in the DEGAS2 routine is determined from the uxes of XGC0 ions to the material surfaces. The kinetic-kinetic plasma-neutral transport capability is demonstrated with example pedestal fueling simulations.

  10. A Parallel Monte Carlo Code for Simulating Collisional N-body Systems

    CERN Document Server

    Pattabiraman, Bharath; Liao, Wei-Keng; Choudhary, Alok; Kalogera, Vassiliki; Memik, Gokhan; Rasio, Frederic A

    2012-01-01

    We present a new parallel code for computing the dynamical evolution of collisional N-body systems with up to N~10^7 particles. Our code is based on the the H\\'enon Monte Carlo method for solving the Fokker-Planck equation, and makes assumptions of spherical symmetry and dynamical equilibrium. The principal algorithmic developments involve optimizing data structures, and the introduction of a parallel random number generation scheme, as well as a parallel sorting algorithm, required to find nearest neighbors for interactions and to compute the gravitational potential. The new algorithms we introduce along with our choice of decomposition scheme minimize communication costs and ensure optimal distribution of data and workload among the processing units. The implementation uses the Message Passing Interface (MPI) library for communication, which makes it portable to many different supercomputing architectures. We validate the code by calculating the evolution of clusters with initial Plummer distribution functi...

  11. Comparison of Geant4-DNA simulation of S-values with other Monte Carlo codes

    Energy Technology Data Exchange (ETDEWEB)

    André, T. [Université Bordeaux 1, CENBG, UMR 5797, F-33170 Gradignan (France); CNRS, IN2P3, CENBG, UMR 5797, F-33170 Gradignan (France); Morini, F. [Research Group of Theoretical Chemistry and Molecular Modelling, Hasselt University, Agoralaan Gebouw D, B-3590 Diepenbeek (Belgium); Karamitros, M. [Université Bordeaux 1, CENBG, UMR 5797, F-33170 Gradignan (France); CNRS, IN2P3, CENBG, UMR 5797, F-33170 Gradignan (France); CNRS, INCIA, UMR 5287, F-33400 Talence (France); Delorme, R. [LPSC, Université Joseph Fourier Grenoble 1, CNRS/IN2P3, Grenoble INP, 38026 Grenoble (France); CEA, LIST, F-91191 Gif-sur-Yvette (France); Le Loirec, C. [CEA, LIST, F-91191 Gif-sur-Yvette (France); Campos, L. [Departamento de Física, Universidade Federal de Sergipe, São Cristóvão (Brazil); Champion, C. [Université Bordeaux 1, CENBG, UMR 5797, F-33170 Gradignan (France); CNRS, IN2P3, CENBG, UMR 5797, F-33170 Gradignan (France); Groetz, J.-E.; Fromm, M. [Université de Franche-Comté, Laboratoire Chrono-Environnement, UMR CNRS 6249, Besançon (France); Bordage, M.-C. [Laboratoire Plasmas et Conversion d’Énergie, UMR 5213 CNRS-INPT-UPS, Université Paul Sabatier, Toulouse (France); Perrot, Y. [Laboratoire de Physique Corpusculaire, UMR 6533, Aubière (France); Barberet, Ph. [Université Bordeaux 1, CENBG, UMR 5797, F-33170 Gradignan (France); CNRS, IN2P3, CENBG, UMR 5797, F-33170 Gradignan (France); and others

    2014-01-15

    Monte Carlo simulations of S-values have been carried out with the Geant4-DNA extension of the Geant4 toolkit. The S-values have been simulated for monoenergetic electrons with energies ranging from 0.1 keV up to 20 keV, in liquid water spheres (for four radii, chosen between 10 nm and 1 μm), and for electrons emitted by five isotopes of iodine (131, 132, 133, 134 and 135), in liquid water spheres of varying radius (from 15 μm up to 250 μm). The results have been compared to those obtained from other Monte Carlo codes and from other published data. The use of the Kolmogorov–Smirnov test has allowed confirming the statistical compatibility of all simulation results.

  12. Selected organ dose conversion coefficients for external photons calculated using ICRP adult voxel phantoms and Monte Carlo code FLUKA.

    Science.gov (United States)

    Patni, H K; Nadar, M Y; Akar, D K; Bhati, S; Sarkar, P K

    2011-11-01

    The adult reference male and female computational voxel phantoms recommended by ICRP are adapted into the Monte Carlo transport code FLUKA. The FLUKA code is then utilised for computation of dose conversion coefficients (DCCs) expressed in absorbed dose per air kerma free-in-air for colon, lungs, stomach wall, breast, gonads, urinary bladder, oesophagus, liver and thyroid due to a broad parallel beam of mono-energetic photons impinging in anterior-posterior and posterior-anterior directions in the energy range of 15 keV-10 MeV. The computed DCCs of colon, lungs, stomach wall and breast are found to be in good agreement with the results published in ICRP publication 110. The present work thus validates the use of FLUKA code in computation of organ DCCs for photons using ICRP adult voxel phantoms. Further, the DCCs for gonads, urinary bladder, oesophagus, liver and thyroid are evaluated and compared with results published in ICRP 74 in the above-mentioned energy range and geometries. Significant differences in DCCs are observed for breast, testis and thyroid above 1 MeV, and for most of the organs at energies below 60 keV in comparison with the results published in ICRP 74. The DCCs of female voxel phantom were found to be higher in comparison with male phantom for almost all organs in both the geometries.

  13. Progress Towards Optimally Efficient Schemes for Monte Carlo Thermal Radiation Transport

    Energy Technology Data Exchange (ETDEWEB)

    Smedley-Stevenson, R P; Brooks III, E D

    2007-09-26

    In this summary we review the complementary research being undertaken at AWE and LLNL aimed at developing optimally efficient algorithms for Monte Carlo thermal radiation transport based on the difference formulation. We conclude by presenting preliminary results on the application of Newton-Krylov methods for solving the Symbolic Implicit Monte Carlo (SIMC) energy equation.

  14. Quality control of the new Monte Carlo code Multi plan for Cyberknife planner; Control de calidad del nuevo codigo Monte Carlo de planificador multiplan para Cyberknife

    Energy Technology Data Exchange (ETDEWEB)

    Fayos Ferrer, F.; Antolin Sanmartin, E.; Simon de Blas, R.; Palazon Cano, I.; Bertomeu Padin, T.; Gutierrez Sarraga, J.; Rey Portoles, G.

    2011-07-01

    This paper is subjected to various tests including Monte Carlo dosimetric the code in the latest versions of Multi plan Accuracy planner. They compare their results and Ray-Tracing Algorithm (RT), present from the earliest versions, with the experimental results obtained by photographic dosimetry and ionization chamber measurements.

  15. 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.

  16. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech., Seoul (Korea, Republic of); Choi, Hoon; Ha, Sang Jun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code.

  18. Preliminary analyses for HTTR`s start-up physics tests by Monte Carlo code MVP

    Energy Technology Data Exchange (ETDEWEB)

    Nojiri, Naoki [Science and Technology Agency, Tokyo (Japan); Nakano, Masaaki; Ando, Hiroei; Fujimoto, Nozomu; Takeuchi, Mitsuo; Fujisaki, Shingo; Yamashita, Kiyonobu

    1998-08-01

    Analyses of start-up physics tests for High Temperature Engineering Test Reactor (HTTR) have been carried out by Monte Carlo code MVP based on continuous energy method. Heterogeneous core structures were modified precisely, such as the fuel compacts, fuel rods, coolant channels, burnable poisons, control rods, control rod insertion holes, reserved shutdown pellet insertion holes, gaps between graphite blocks, etc. Such precise modification of the core structures was difficult with diffusion calculation. From the analytical results, the followings were confirmed; The first criticality will be achieved around 16 fuel columns loaded. The reactivity at the first criticality can be controlled by only one control rod located at the center of the core with other fifteen control rods fully withdrawn. The excess reactivity, reactor shutdown margin and control rod criticality positions have been evaluated. These results were used for planning of the start-up physics tests. This report presents analyses of start-up physics tests for HTTR by MVP code. (author)

  19. Generation of XS library for the reflector of VVER reactor core using Monte Carlo code Serpent

    Science.gov (United States)

    Usheva, K. I.; Kuten, S. A.; Khruschinsky, A. A.; Babichev, L. F.

    2017-01-01

    A physical model of the radial and axial reflector of VVER-1200-like reactor core has been developed. Five types of radial reflector with different material composition exist for the VVER reactor core and 1D and 2D models were developed for all of them. Axial top and bottom reflectors are described by the 1D model. A two-group XS library for diffusion code DYN3D has been generated for all types of reflectors by using Serpent 2 Monte Carlo code. Power distribution in the reactor core calculated in DYN3D is flattened in the core central region to more extent in the 2D model of the radial reflector than in its 1D model.

  20. Coded aperture coherent scatter imaging for breast cancer detection: a Monte Carlo evaluation

    Science.gov (United States)

    Lakshmanan, Manu N.; Morris, Robert E.; Greenberg, Joel A.; Samei, Ehsan; Kapadia, Anuj J.

    2016-03-01

    It is known that conventional x-ray imaging provides a maximum contrast between cancerous and healthy fibroglandular breast tissues of 3% based on their linear x-ray attenuation coefficients at 17.5 keV, whereas coherent scatter signal provides a maximum contrast of 19% based on their differential coherent scatter cross sections. Therefore in order to exploit this potential contrast, we seek to evaluate the performance of a coded- aperture coherent scatter imaging system for breast cancer detection and investigate its accuracy using Monte Carlo simulations. In the simulations we modeled our experimental system, which consists of a raster-scanned pencil beam of x-rays, a bismuth-tin coded aperture mask comprised of a repeating slit pattern with 2-mm periodicity, and a linear-array of 128 detector pixels with 6.5-keV energy resolution. The breast tissue that was scanned comprised a 3-cm sample taken from a patient-based XCAT breast phantom containing a tomosynthesis- based realistic simulated lesion. The differential coherent scatter cross section was reconstructed at each pixel in the image using an iterative reconstruction algorithm. Each pixel in the reconstructed image was then classified as being either air or the type of breast tissue with which its normalized reconstructed differential coherent scatter cross section had the highest correlation coefficient. Comparison of the final tissue classification results with the ground truth image showed that the coded aperture imaging technique has a cancerous pixel detection sensitivity (correct identification of cancerous pixels), specificity (correctly ruling out healthy pixels as not being cancer) and accuracy of 92.4%, 91.9% and 92.0%, respectively. Our Monte Carlo evaluation of our experimental coded aperture coherent scatter imaging system shows that it is able to exploit the greater contrast available from coherently scattered x-rays to increase the accuracy of detecting cancerous regions within the breast.

  1. Spread-out Bragg peak and monitor units calculation with the Monte Carlo code MCNPX.

    Science.gov (United States)

    Hérault, J; Iborra, N; Serrano, B; Chauvel, P

    2007-02-01

    The aim of this work was to study the dosimetric potential of the Monte Carlo code MCNPX applied to the protontherapy field. For series of clinical configurations a comparison between simulated and experimental data was carried out, using the proton beam line of the MEDICYC isochronous cyclotron installed in the Centre Antoine Lacassagne in Nice. The dosimetric quantities tested were depth-dose distributions, output factors, and monitor units. For each parameter, the simulation reproduced accurately the experiment, which attests the quality of the choices made both in the geometrical description and in the physics parameters for beam definition. These encouraging results enable us today to consider a simplification of quality control measurements in the future. Monitor Units calculation is planned to be carried out with preestablished Monte Carlo simulation data. The measurement, which was until now our main patient dose calibration system, will be progressively replaced by computation based on the MCNPX code. This determination of Monitor Units will be controlled by an independent semi-empirical calculation.

  2. MONTE CARLO NEUTRINO TRANSPORT THROUGH REMNANT DISKS FROM NEUTRON STAR MERGERS

    Energy Technology Data Exchange (ETDEWEB)

    Richers, Sherwood; Ott, Christian D. [TAPIR, Mailcode 350-17, Walter Burke Institute for Theoretical Physics, California Institute of Technology, Pasadena, CA 91125 (United States); Kasen, Daniel; Fernández, Rodrigo [Department of Astronomy and Theoretical Astrophysics Center, University of California, Berkeley, CA 94720 (United States); O’Connor, Evan [Department of Physics, Campus Code 8202, North Carolina State University, Raleigh, NC 27695 (United States)

    2015-11-01

    We present Sedonu, a new open source, steady-state, special relativistic Monte Carlo (MC) neutrino transport code, available at bitbucket.org/srichers/sedonu. The code calculates the energy- and angle-dependent neutrino distribution function on fluid backgrounds of any number of spatial dimensions, calculates the rates of change of fluid internal energy and electron fraction, and solves for the equilibrium fluid temperature and electron fraction. We apply this method to snapshots from two-dimensional simulations of accretion disks left behind by binary neutron star mergers, varying the input physics and comparing to the results obtained with a leakage scheme for the cases of a central black hole and a central hypermassive neutron star. Neutrinos are guided away from the densest regions of the disk and escape preferentially around 45° from the equatorial plane. Neutrino heating is strengthened by MC transport a few scale heights above the disk midplane near the innermost stable circular orbit, potentially leading to a stronger neutrino-driven wind. Neutrino cooling in the dense midplane of the disk is stronger when using MC transport, leading to a globally higher cooling rate by a factor of a few and a larger leptonization rate by an order of magnitude. We calculate neutrino pair annihilation rates and estimate that an energy of 2.8 × 10{sup 46} erg is deposited within 45° of the symmetry axis over 300 ms when a central BH is present. Similarly, 1.9 × 10{sup 48} erg is deposited over 3 s when an HMNS sits at the center, but neither estimate is likely to be sufficient to drive a gamma-ray burst jet.

  3. DETERMINISTIC TRANSPORT METHODS AND CODES AT LOS ALAMOS

    Energy Technology Data Exchange (ETDEWEB)

    J. E. MOREL

    1999-06-01

    The purposes of this paper are to: Present a brief history of deterministic transport methods development at Los Alamos National Laboratory from the 1950's to the present; Discuss the current status and capabilities of deterministic transport codes at Los Alamos; and Discuss future transport needs and possible future research directions. Our discussion of methods research necessarily includes only a small fraction of the total research actually done. The works that have been included represent a very subjective choice on the part of the author that was strongly influenced by his personal knowledge and experience. The remainder of this paper is organized in four sections: the first relates to deterministic methods research performed at Los Alamos, the second relates to production codes developed at Los Alamos, the third relates to the current status of transport codes at Los Alamos, and the fourth relates to future research directions at Los Alamos.

  4. SU-E-T-578: MCEBRT, A Monte Carlo Code for External Beam Treatment Plan Verifications

    Energy Technology Data Exchange (ETDEWEB)

    Chibani, O; Ma, C [Fox Chase Cancer Center, Philadelphia, PA (United States); Eldib, A [Fox Chase Cancer Center, Philadelphia, PA (United States); Al-Azhar University, Cairo (Egypt)

    2014-06-01

    Purpose: Present a new Monte Carlo code (MCEBRT) for patient-specific dose calculations in external beam radiotherapy. The code MLC model is benchmarked and real patient plans are re-calculated using MCEBRT and compared with commercial TPS. Methods: MCEBRT is based on the GEPTS system (Med. Phys. 29 (2002) 835–846). Phase space data generated for Varian linac photon beams (6 – 15 MV) are used as source term. MCEBRT uses a realistic MLC model (tongue and groove, rounded ends). Patient CT and DICOM RT files are used to generate a 3D patient phantom and simulate the treatment configuration (gantry, collimator and couch angles; jaw positions; MLC sequences; MUs). MCEBRT dose distributions and DVHs are compared with those from TPS in absolute way (Gy). Results: Calculations based on the developed MLC model closely matches transmission measurements (pin-point ionization chamber at selected positions and film for lateral dose profile). See Fig.1. Dose calculations for two clinical cases (whole brain irradiation with opposed beams and lung case with eight fields) are carried out and outcomes are compared with the Eclipse AAA algorithm. Good agreement is observed for the brain case (Figs 2-3) except at the surface where MCEBRT dose can be higher by 20%. This is due to better modeling of electron contamination by MCEBRT. For the lung case an overall good agreement (91% gamma index passing rate with 3%/3mm DTA criterion) is observed (Fig.4) but dose in lung can be over-estimated by up to 10% by AAA (Fig.5). CTV and PTV DVHs from TPS and MCEBRT are nevertheless close (Fig.6). Conclusion: A new Monte Carlo code is developed for plan verification. Contrary to phantombased QA measurements, MCEBRT simulate the exact patient geometry and tissue composition. MCEBRT can be used as extra verification layer for plans where surface dose and tissue heterogeneity are an issue.

  5. Investigation of Nuclear Data Libraries with TRIPOLI-4 Monte Carlo Code for Sodium-cooled Fast Reactors

    Science.gov (United States)

    Lee, Y.-K.; Brun, E.

    2014-04-01

    The Sodium-cooled fast neutron reactor ASTRID is currently under design and development in France. Traditional ECCO/ERANOS fast reactor code system used for ASTRID core design calculations relies on multi-group JEFF-3.1.1 data library. To gauge the use of ENDF/B-VII.0 and JEFF-3.1.1 nuclear data libraries in the fast reactor applications, two recent OECD/NEA computational benchmarks specified by Argonne National Laboratory were calculated. Using the continuous-energy TRIPOLI-4 Monte Carlo transport code, both ABR-1000 MWth MOX core and metallic (U-Pu) core were investigated. Under two different fast neutron spectra and two data libraries, ENDF/B-VII.0 and JEFF-3.1.1, reactivity impact studies were performed. Using JEFF-3.1.1 library under the BOEC (Beginning of equilibrium cycle) condition, high reactivity effects of 808 ± 17 pcm and 1208 ± 17 pcm were observed for ABR-1000 MOX core and metallic core respectively. To analyze the causes of these differences in reactivity, several TRIPOLI-4 runs using mixed data libraries feature allow us to identify the nuclides and the nuclear data accounting for the major part of the observed reactivity discrepancies.

  6. Modeling of Anomalous Transport in Tokamaks with FACETS code

    Science.gov (United States)

    Pankin, A. Y.; Batemann, G.; Kritz, A.; Rafiq, T.; Vadlamani, S.; Hakim, A.; Kruger, S.; Miah, M.; Rognlien, T.

    2009-05-01

    The FACETS code, a whole-device integrated modeling code that self-consistently computes plasma profiles for the plasma core and edge in tokamaks, has been recently developed as a part of the SciDAC project for core-edge simulations. A choice of transport models is available in FACETS through the FMCFM interface [1]. Transport models included in FMCFM have specific ranges of applicability, which can limit their use to parts of the plasma. In particular, the GLF23 transport model does not include the resistive ballooning effects that can be important in the tokamak pedestal region and GLF23 typically under-predicts the anomalous fluxes near the magnetic axis [2]. The TGLF and GYRO transport models have similar limitations [3]. A combination of transport models that covers the entire discharge domain is studied using FACETS in a realistic tokamak geometry. Effective diffusivities computed with the FMCFM transport models are extended to the region near the separatrix to be used in the UEDGE code within FACETS. 1. S. Vadlamani et al. (2009) %First time-dependent transport simulations using GYRO and NCLASS within FACETS (this meeting).2. T. Rafiq et al. (2009) %Simulation of electron thermal transport in H-mode discharges Submitted to Phys. Plasmas.3. C. Holland et al. (2008) %Validation of gyrokinetic transport simulations using %DIII-D core turbulence measurements Proc. of IAEA FEC (Switzerland, 2008)

  7. Mesh-based Monte Carlo code for fluorescence modeling in complex tissues with irregular boundaries

    Science.gov (United States)

    Wilson, Robert H.; Chen, Leng-Chun; Lloyd, William; Kuo, Shiuhyang; Marcelo, Cynthia; Feinberg, Stephen E.; Mycek, Mary-Ann

    2011-07-01

    There is a growing need for the development of computational models that can account for complex tissue morphology in simulations of photon propagation. We describe the development and validation of a user-friendly, MATLAB-based Monte Carlo code that uses analytically-defined surface meshes to model heterogeneous tissue geometry. The code can use information from non-linear optical microscopy images to discriminate the fluorescence photons (from endogenous or exogenous fluorophores) detected from different layers of complex turbid media. We present a specific application of modeling a layered human tissue-engineered construct (Ex Vivo Produced Oral Mucosa Equivalent, EVPOME) designed for use in repair of oral tissue following surgery. Second-harmonic generation microscopic imaging of an EVPOME construct (oral keratinocytes atop a scaffold coated with human type IV collagen) was employed to determine an approximate analytical expression for the complex shape of the interface between the two layers. This expression can then be inserted into the code to correct the simulated fluorescence for the effect of the irregular tissue geometry.

  8. X-ray simulation with the Monte Carlo code PENELOPE. Application to Quality Control.

    Science.gov (United States)

    Pozuelo, F; Gallardo, S; Querol, A; Verdú, G; Ródenas, J

    2012-01-01

    A realistic knowledge of the energy spectrum is very important in Quality Control (QC) of X-ray tubes in order to reduce dose to patients. However, due to the implicit difficulties to measure the X-ray spectrum accurately, it is not normally obtained in routine QC. Instead, some parameters are measured and/or calculated. PENELOPE and MCNP5 codes, based on the Monte Carlo method, can be used as complementary tools to verify parameters measured in QC. These codes allow estimating Bremsstrahlung and characteristic lines from the anode taking into account specific characteristics of equipment. They have been applied to simulate an X-ray spectrum. Results are compared with theoretical IPEM 78 spectrum. A sensitivity analysis has been developed to estimate the influence on simulated spectra of important parameters used in simulation codes. With this analysis it has been obtained that the FORCE factor is the most important parameter in PENELOPE simulations. FORCE factor, which is a variance reduction method, improves the simulation but produces hard increases of computer time. The value of FORCE should be optimized so that a good agreement of simulated and theoretical spectra is reached, but with a reduction of computer time. Quality parameters such as Half Value Layer (HVL) can be obtained with the PENELOPE model developed, but FORCE takes such a high value that computer time is hardly increased. On the other hand, depth dose assessment can be achieved with acceptable results for small values of FORCE.

  9. Radiation Transport for Explosive Outflows: A Multigroup Hybrid Monte Carlo Method

    CERN Document Server

    Wollaeger, Ryan T; Graziani, Carlo; Couch, Sean M; Jordan, George C; Lamb, Donald Q; Moses, Gregory A

    2013-01-01

    We explore the application of Implicit Monte Carlo (IMC) and Discrete Diffusion Monte Carlo (DDMC) to radiation transport in strong fluid outflows with structured opacity. The IMC method of Fleck & Cummings is a stochastic computational technique for nonlinear radiation transport. IMC is partially implicit in time and may suffer in efficiency when tracking Monte Carlo particles through optically thick materials. The DDMC method of Densmore accelerates an IMC computation where the domain is diffusive. Recently, Abdikamalov extended IMC and DDMC to multigroup, velocity-dependent neutrino transport with the intent of modeling neutrino dynamics in core-collapse supernovae. Densmore has also formulated a multifrequency extension to the originally grey DDMC method. In this article we rigorously formulate IMC and DDMC over a high-velocity Lagrangian grid for possible application to photon transport in the post-explosion phase of Type Ia supernovae. The method described is suitable for a large variety of non-mono...

  10. Calculation of electron and isotopes dose point kernels with fluka Monte Carlo code for dosimetry in nuclear medicine therapy

    Energy Technology Data Exchange (ETDEWEB)

    Botta, F; Di Dia, A; Pedroli, G; Mairani, A; Battistoni, G; Fasso, A; Ferrari, A; Ferrari, M; Paganelli, G

    2011-06-01

    The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, fluka Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, fluka has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one.Methods: fluka DPKs have been calculated in both water and compact bone for monoenergetic electrons (10–3 MeV) and for beta emitting isotopes commonly used for therapy (89Sr, 90Y, 131I, 153Sm, 177Lu, 186Re, and 188Re). Point isotropic sources have been simulated at the center of a water (bone) sphere, and deposed energy has been tallied in concentric shells. fluka outcomes have been compared to penelope v.2008 results, calculated in this study as well. Moreover, in case of monoenergetic electrons in water, comparison with the data from the literature (etran, geant4, mcnpx) has been done. Maximum percentage differences within 0.8·RCSDA and 0.9·RCSDA for monoenergetic electrons (RCSDA being the continuous slowing down approximation range) and within 0.8·X90 and 0.9·X90 for isotopes (X90 being the radius of the sphere in which 90% of the emitted energy is absorbed) have been computed, together with the average percentage difference within 0.9·RCSDA and 0.9·X90 for electrons and isotopes, respectively.Results: Concerning monoenergetic electrons, within 0.8·RCSDA (where 90%–97% of the particle energy is deposed), fluka and penelope agree mostly within 7%, except for 10 and 20 keV electrons (12% in water, 8

  11. Calculation of electron and isotopes dose point kernels with fluka Monte Carlo code for dosimetry in nuclear medicine therapy

    Energy Technology Data Exchange (ETDEWEB)

    Botta, F.; Mairani, A.; Battistoni, G.; Cremonesi, M.; Di Dia, A.; Fasso, A.; Ferrari, A.; Ferrari, M.; Paganelli, G.; Pedroli, G.; Valente, M. [Medical Physics Department, European Institute of Oncology, Via Ripamonti 435, 20141 Milan (Italy); Istituto Nazionale di Fisica Nucleare (I.N.F.N.), Via Celoria 16, 20133 Milan (Italy); Medical Physics Department, European Institute of Oncology, Via Ripamonti 435, 20141 Milan (Italy); Jefferson Lab, 12000 Jefferson Avenue, Newport News, Virginia 23606 (United States); CERN, 1211 Geneva 23 (Switzerland); Medical Physics Department, European Institute of Oncology, Milan (Italy); Nuclear Medicine Department, European Institute of Oncology, Via Ripamonti 435, 2014 Milan (Italy); Medical Physics Department, European Institute of Oncology, Via Ripamonti 435, 20141 Milan (Italy); FaMAF, Universidad Nacional de Cordoba and CONICET, Cordoba, Argentina C.P. 5000 (Argentina)

    2011-07-15

    Purpose: The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, fluka Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, fluka has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one. Methods: fluka DPKs have been calculated in both water and compact bone for monoenergetic electrons (10{sup -3} MeV) and for beta emitting isotopes commonly used for therapy ({sup 89}Sr, {sup 90}Y, {sup 131}I, {sup 153}Sm, {sup 177}Lu, {sup 186}Re, and {sup 188}Re). Point isotropic sources have been simulated at the center of a water (bone) sphere, and deposed energy has been tallied in concentric shells. fluka outcomes have been compared to penelope v.2008 results, calculated in this study as well. Moreover, in case of monoenergetic electrons in water, comparison with the data from the literature (etran, geant4, mcnpx) has been done. Maximum percentage differences within 0.8{center_dot}R{sub CSDA} and 0.9{center_dot}R{sub CSDA} for monoenergetic electrons (R{sub CSDA} being the continuous slowing down approximation range) and within 0.8{center_dot}X{sub 90} and 0.9{center_dot}X{sub 90} for isotopes (X{sub 90} being the radius of the sphere in which 90% of the emitted energy is absorbed) have been computed, together with the average percentage difference within 0.9{center_dot}R{sub CSDA} and 0.9{center_dot}X{sub 90} for electrons and isotopes, respectively. Results: Concerning monoenergetic electrons

  12. The FLUKA code for application of Monte Carlo methods to promote high precision ion beam therapy

    CERN Document Server

    Parodi, K; Cerutti, F; Ferrari, A; Mairani, A; Paganetti, H; Sommerer, F

    2010-01-01

    Monte Carlo (MC) methods are increasingly being utilized to support several aspects of commissioning and clinical operation of ion beam therapy facilities. In this contribution two emerging areas of MC applications are outlined. The value of MC modeling to promote accurate treatment planning is addressed via examples of application of the FLUKA code to proton and carbon ion therapy at the Heidelberg Ion Beam Therapy Center in Heidelberg, Germany, and at the Proton Therapy Center of Massachusetts General Hospital (MGH) Boston, USA. These include generation of basic data for input into the treatment planning system (TPS) and validation of the TPS analytical pencil-beam dose computations. Moreover, we review the implementation of PET/CT (Positron-Emission-Tomography / Computed- Tomography) imaging for in-vivo verification of proton therapy at MGH. Here, MC is used to calculate irradiation-induced positron-emitter production in tissue for comparison with the +-activity measurement in order to infer indirect infor...

  13. Exploring Monte Carlo methods

    CERN Document Server

    Dunn, William L

    2012-01-01

    Exploring Monte Carlo Methods is a basic text that describes the numerical methods that have come to be known as "Monte Carlo." The book treats the subject generically through the first eight chapters and, thus, should be of use to anyone who wants to learn to use Monte Carlo. The next two chapters focus on applications in nuclear engineering, which are illustrative of uses in other fields. Five appendices are included, which provide useful information on probability distributions, general-purpose Monte Carlo codes for radiation transport, and other matters. The famous "Buffon's needle proble

  14. Event-by-event Monte Carlo simulation of radiation transport in vapor and liquid water

    Science.gov (United States)

    Papamichael, Georgios Ioannis

    A Monte-Carlo Simulation is presented for Radiation Transport in water. This process is of utmost importance, having applications in oncology and therapy of cancer, in protecting people and the environment, waste management, radiation chemistry and on some solid-state detectors. It's also a phenomenon of interest in microelectronics on satellites in orbit that are subject to the solar radiation and in space-craft design for deep-space missions receiving background radiation. The interaction of charged particles with the medium is primarily due to their electromagnetic field. Three types of interaction events are considered: Elastic scattering, impact excitation and impact ionization. Secondary particles (electrons) can be generated by ionization. At each stage, along with the primary particle we explicitly follow all secondary electrons (and subsequent generations). Theoretical, semi-empirical and experimental formulae with suitable corrections have been used in each case to model the cross sections governing the quantum mechanical process of interactions, thus determining stochastically the energy and direction of outgoing particles following an event. Monte-Carlo sampling techniques have been applied to accurate probability distribution functions describing the primary particle track and all secondary particle-medium interaction. A simple account of the simulation code and a critical exposition of its underlying assumptions (often missing in the relevant literature) are also presented with reference to the model cross sections. Model predictions are in good agreement with existing computational data and experimental results. By relying heavily on a theoretical formulation, instead of merely fitting data, it is hoped that the model will be of value in a wider range of applications. Possible future directions that are the object of further research are pointed out.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-01

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

  16. egs_brachy: a versatile and fast Monte Carlo code for brachytherapy

    Science.gov (United States)

    Chamberland, Marc J. P.; Taylor, Randle E. P.; Rogers, D. W. O.; Thomson, Rowan M.

    2016-12-01

    egs_brachy is a versatile and fast Monte Carlo (MC) code for brachytherapy applications. It is based on the EGSnrc code system, enabling simulation of photons and electrons. Complex geometries are modelled using the EGSnrc C++ class library and egs_brachy includes a library of geometry models for many brachytherapy sources, in addition to eye plaques and applicators. Several simulation efficiency enhancing features are implemented in the code. egs_brachy is benchmarked by comparing TG-43 source parameters of three source models to previously published values. 3D dose distributions calculated with egs_brachy are also compared to ones obtained with the BrachyDose code. Well-defined simulations are used to characterize the effectiveness of many efficiency improving techniques, both as an indication of the usefulness of each technique and to find optimal strategies. Efficiencies and calculation times are characterized through single source simulations and simulations of idealized and typical treatments using various efficiency improving techniques. In general, egs_brachy shows agreement within uncertainties with previously published TG-43 source parameter values. 3D dose distributions from egs_brachy and BrachyDose agree at the sub-percent level. Efficiencies vary with radionuclide and source type, number of sources, phantom media, and voxel size. The combined effects of efficiency-improving techniques in egs_brachy lead to short calculation times: simulations approximating prostate and breast permanent implant (both with (2 mm)3 voxels) and eye plaque (with (1 mm)3 voxels) treatments take between 13 and 39 s, on a single 2.5 GHz Intel Xeon E5-2680 v3 processor core, to achieve 2% average statistical uncertainty on doses within the PTV. egs_brachy will be released as free and open source software to the research community.

  17. Simulation of a nuclear densimeter using the Monte Carlo MCNP-4C code; Simulacao de um densimetro nuclear utilizando o codigo Monte Carlo MCNP-4C

    Energy Technology Data Exchange (ETDEWEB)

    Penna, Rodrigo [UNI-BH, Belo Horizonte, MG (Brazil). Dept. de Ciencias Biologicas, Ambientais e da Saude (DCBAS/DCET); Silva, Clemente Jose Gusmao Carneiro da [Universidade Estadual de Santa Cruz, UESC, Ilheus, BA (Brazil); Gomes, Paulo Mauricio Costa [Universidade FUMEC, Belo Horizonte, MG (Brazil)

    2008-07-01

    Viability of building a nuclear wood densimeter based on low energy photons Compton scattering was done using Monte Carlo code (MCNP- 4C). It is simulated a collimated 60 keV beam of gamma rays emitted by {sup 241}Am source reaching wood blocks. Backscattered radiation by these blocks was calculated. Photons scattered were correlated with blocks of different wood densities. Results showed a linear relationship on wood density and scattered photons, therefore the viability of this wood densimeter. (author)

  18. DANTSYS: A diffusion accelerated neutral particle transport code system

    Energy Technology Data Exchange (ETDEWEB)

    Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Marr, D.R.; O`Dell, R.D.; Walters, W.F.

    1995-06-01

    The DANTSYS code package includes the following transport codes: ONEDANT, TWODANT, TWODANT/GQ, TWOHEX, and THREEDANT. The DANTSYS code package is a modular computer program package designed to solve the time-independent, multigroup discrete ordinates form of the boltzmann transport equation in several different geometries. The modular construction of the package separates the input processing, the transport equation solving, and the post processing (or edit) functions into distinct code modules: the Input Module, one or more Solver Modules, and the Edit Module, respectively. The Input and Edit Modules are very general in nature and are common to all the Solver Modules. The ONEDANT Solver Module contains a one-dimensional (slab, cylinder, and sphere), time-independent transport equation solver using the standard diamond-differencing method for space/angle discretization. Also included in the package are solver Modules named TWODANT, TWODANT/GQ, THREEDANT, and TWOHEX. The TWODANT Solver Module solves the time-independent two-dimensional transport equation using the diamond-differencing method for space/angle discretization. The authors have also introduced an adaptive weighted diamond differencing (AWDD) method for the spatial and angular discretization into TWODANT as an option. The TWOHEX Solver Module solves the time-independent two-dimensional transport equation on an equilateral triangle spatial mesh. The THREEDANT Solver Module solves the time independent, three-dimensional transport equation for XYZ and RZ{Theta} symmetries using both diamond differencing with set-to-zero fixup and the AWDD method. The TWODANT/GQ Solver Module solves the 2-D transport equation in XY and RZ symmetries using a spatial mesh of arbitrary quadrilaterals. The spatial differencing method is based upon the diamond differencing method with set-to-zero fixup with changes to accommodate the generalized spatial meshing.

  19. New electron multiple scattering distributions for Monte Carlo transport simulation

    Energy Technology Data Exchange (ETDEWEB)

    Chibani, Omar (Haut Commissariat a la Recherche (C.R.S.), 2 Boulevard Franz Fanon, Alger B.P. 1017, Alger-Gare (Algeria)); Patau, Jean Paul (Laboratoire de Biophysique et Biomathematiques, Faculte des Sciences Pharmaceutiques, Universite Paul Sabatier, 35 Chemin des Maraichers, 31062 Toulouse cedex (France))

    1994-10-01

    New forms of electron (positron) multiple scattering distributions are proposed. The first is intended for use in the conditions of validity of the Moliere theory. The second distribution takes place when the electron path is so short that only few elastic collisions occur. These distributions are adjustable formulas. The introduction of some parameters allows impositions of the correct value of the first moment. Only positive and analytic functions were used in constructing the present expressions. This makes sampling procedures easier. Systematic tests are presented and some Monte Carlo simulations, as benchmarks, are carried out. ((orig.))

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

    Science.gov (United States)

    Nelson, W. R.; Namito, Yoshihito

    1990-03-01

    In this paper we present an overview of the EGS4 Code System -- a general purpose package for the Monte Carlo simulation of the transport of electrons and photons. During the last 10-15 years EGS has been widely used to design accelerators and detectors for high-energy physics. More recently the code has been found to be of tremendous use in medical radiation physics and dosimetry. The problem-solving capabilities of EGS4 will be demonstrated by means of a variety of practical examples. To facilitate this review, we will take advantage of a new add-on package, called SHOWGRAF, to display particle trajectories in complicated geometries. These are shown as 2-D laser pictures in the written paper and as photographic slides of a 3-D high-resolution color monitor during the oral presentation. 11 refs., 15 figs.

  1. Monte Carlo model of neutral-particle transport in diverted plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Heifetz, D.; Post, D.; Petravic, M.; Weisheit, J.; Bateman, G.

    1981-11-01

    The transport of neutral atoms and molecules in the edge and divertor regions of fusion experiments has been calculated using Monte-Carlo techniques. The deuterium, tritium, and helium atoms are produced by recombination in the plasma and at the walls. The relevant collision processes of charge exchange, ionization, and dissociation between the neutrals and the flowing plasma electrons and ions are included, along with wall reflection models. General two-dimensional wall and plasma geometries are treated in a flexible manner so that varied configurations can be easily studied. The algorithm uses a pseudo-collision method. Splitting with Russian roulette, suppression of absorption, and efficient scoring techniques are used to reduce the variance. The resulting code is sufficiently fast and compact to be incorporated into iterative treatments of plasma dynamics requiring numerous neutral profiles. The calculation yields the neutral gas densities, pressures, fluxes, ionization rates, momentum transfer rates, energy transfer rates, and wall sputtering rates. Applications have included modeling of proposed INTOR/FED poloidal divertor designs and other experimental devices.

  2. Monte Carlo simulation of radiation transport in human skin with rigorous treatment of curved tissue boundaries.

    Science.gov (United States)

    Majaron, Boris; Milanič, Matija; Premru, Jan

    2015-01-01

    In three-dimensional (3-D) modeling of light transport in heterogeneous biological structures using the Monte Carlo (MC) approach, space is commonly discretized into optically homogeneous voxels by a rectangular spatial grid. Any round or oblique boundaries between neighboring tissues thus become serrated, which raises legitimate concerns about the realism of modeling results with regard to reflection and refraction of light on such boundaries. We analyze the related effects by systematic comparison with an augmented 3-D MC code, in which analytically defined tissue boundaries are treated in a rigorous manner. At specific locations within our test geometries, energy deposition predicted by the two models can vary by 10%. Even highly relevant integral quantities, such as linear density of the energy absorbed by modeled blood vessels, differ by up to 30%. Most notably, the values predicted by the customary model vary strongly and quite erratically with the spatial discretization step and upon minor repositioning of the computational grid. Meanwhile, the augmented model shows no such unphysical behavior. Artifacts of the former approach do not converge toward zero with ever finer spatial discretization, confirming that it suffers from inherent deficiencies due to inaccurate treatment of reflection and refraction at round tissue boundaries.

  3. Differential Cross Section Kinematics for 3-dimensional Transport Codes

    Science.gov (United States)

    Norbury, John W.; Dick, Frank

    2008-01-01

    In support of the development of 3-dimensional transport codes, this paper derives the relevant relativistic particle kinematic theory. Formulas are given for invariant, spectral and angular distributions in both the lab (spacecraft) and center of momentum frames, for collisions involving 2, 3 and n - body final states.

  4. Monte Carlo study of electron transport in monolayer silicene

    Science.gov (United States)

    Borowik, Piotr; Thobel, Jean-Luc; Adamowicz, Leszek

    2016-11-01

    Electron mobility and diffusion coefficients in monolayer silicene are calculated by Monte Carlo simulations using simplified band structure with linear energy bands. Results demonstrate reasonable agreement with the full-band Monte Carlo method in low applied electric field conditions. Negative differential resistivity is observed and an explanation of the origin of this effect is proposed. Electron mobility and diffusion coefficients are studied in low applied electric field conditions. We demonstrate that a comparison of these parameter values can provide a good check that the calculation is correct. Low-field mobility in silicene exhibits {T}-3 temperature dependence for nondegenerate electron gas conditions and {T}-1 for higher electron concentrations, when degenerate conditions are imposed. It is demonstrated that to explain the relation between mobility and temperature in nondegenerate electron gas the linearity of the band structure has to be taken into account. It is also found that electron-electron scattering only slightly modifies low-field electron mobility in degenerate electron gas conditions.

  5. Dose estimation in space using the Particle and Heavy-Ion Transport code System (PHITS)

    Energy Technology Data Exchange (ETDEWEB)

    Gustafsson, Katarina

    2009-06-15

    The radiation risks in space are well known, but work still needs to be done in order to fully understand the radiation effects on humans and how to minimize the risks especially now when the activity in space is increasing with plans for missions to the Moon and Mars. One goal is to develop transport codes that can estimate the radiation environment and its effects. These would be useful tools for reducing the radiation effects when designing and planning space missions. The Particle and Heavy-Ion Transport code System, PHITS, is a three dimensional Monte Carlo code with great possibilities to perform radiation transport calculations and estimating radiation exposure such as absorbed dose, equivalent dose and dose equivalent. Therefore a benchmarking with experiments performed at the ISS was done and also an estimation of different material's influences on the shielding was made. The simulated results already agree reasonable with the measurements, but can most likely be significantly improved when more realistic shielding geometries will be used. This indicates that PHITS is a useful tool for estimating radiation risks for humans in space and when designing shielding of space crafts

  6. Transport code and nuclear data in intermediate energy region

    Energy Technology Data Exchange (ETDEWEB)

    Hasegawa, Akira; Odama, Naomitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Maekawa, F.; Ueki, K.; Kosaka, K.; Oyama, Y.

    1998-11-01

    We briefly reviewed the problems of intermediate energy nuclear data file and transport codes in connection with processing of the data. This is a summary of our group in the task force on JENDL High Energy File Integral Evaluation (JHEFIE). In this article we stress the necessity of the production of intermediate evaluated nuclear data file up to 3 GeV for the application of accelerator driven transmutation (ADT) system. And also we state the necessity of having our own transport code system to calculate the radiation fields using these evaluated files from the strategic points of view to keep our development of the ADT technology completely free from other conditions outside of our own such as imported codes and data with poor maintenance or unknown accuracy. (author)

  7. Benchmarking of neutron production of heavy-ion transport codes

    Energy Technology Data Exchange (ETDEWEB)

    Remec, I. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6172 (United States); Ronningen, R. M. [Michigan State Univ., National Superconductiong Cyclotron Laboratory, East Lansing, MI 48824-1321 (United States); Heilbronn, L. [Univ. of Tennessee, 1004 Estabrook Rd., Knoxville, TN 37996-2300 (United States)

    2011-07-01

    Document available in abstract form only, full text of document follows: Accurate prediction of radiation fields generated by heavy ion interactions is important in medical applications, space missions, and in design and operation of rare isotope research facilities. In recent years, several well-established computer codes in widespread use for particle and radiation transport calculations have been equipped with the capability to simulate heavy ion transport and interactions. To assess and validate these capabilities, we performed simulations of a series of benchmark-quality heavy ion experiments with the computer codes FLUKA, MARS15, MCNPX, and PHITS. We focus on the comparisons of secondary neutron production. Results are encouraging; however, further improvements in models and codes and additional benchmarking are required. (authors)

  8. MCMini: Monte Carlo on GPGPU

    Energy Technology Data Exchange (ETDEWEB)

    Marcus, Ryan C. [Los Alamos National Laboratory

    2012-07-25

    MCMini is a proof of concept that demonstrates the possibility for Monte Carlo neutron transport using OpenCL with a focus on performance. This implementation, written in C, shows that tracing particles and calculating reactions on a 3D mesh can be done in a highly scalable fashion. These results demonstrate a potential path forward for MCNP or other Monte Carlo codes.

  9. Monte Carlo neutral particle transport through a binary stochastic mixture using chord length sampling

    Science.gov (United States)

    Donovan, Timothy J.

    A Monte Carlo algorithm is developed to estimate the ensemble-averaged behavior of neutral particles within a binary stochastic mixture. A special case stochastic mixture is examined, in which non-overlapping spheres of constant radius are uniformly mixed in a matrix material. Spheres are chosen to represent the stochastic volumes due to their geometric simplicity and because spheres are a common approximation to a large number of applications. The boundaries of the mixture are impenetrable, meaning that spheres in the stochastic mixture cannot be assumed to overlap the mixture boundaries. The algorithm employs a method called Limited Chord Length Sampling (LCLS). While in the matrix material, LCLS uses chord-length sampling to sample the distance to the next stochastic interface. After a surface crossing into a stochastic sphere, transport is treated explicitly until the particle exits or is killed. This capability eliminates the need to explicitly model a representation of the random geometry of the mixture. The algorithm is first proposed and tested against benchmark results for a two dimensional, fixed source model using stand-alone Monte Carlo codes. The algorithm is then implemented and tested in a test version of the Los Alamos M&barbelow;onte C&barbelow;arlo ṉ-p&barbelow;article Code MCNP. This prototype MCNP version has the capability to calculate LCLS results for both fixed source and multiplied source (i.e., eigenvalue) problems. Problems analyzed with MCNP range from simple binary mixtures, designed to test LCLS over a range of optical thicknesses, to a detailed High Temperature Gas Reactor fuel element, which tests the value of LCLS in a current problem of practical significance. Comparisons of LCLS and benchmark results include both accuracy and efficiency comparisons. To ensure conservative efficiency comparisons, the statistical basis for the benchmark technique is derived and a formal method for optimizing the benchmark calculations is developed

  10. Monte Carlo simulation of the electron transport through thin slabs: A comparative study of PENELOPE, GEANT3, GEANT4, EGSnrc and MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Vilches, M. [Servicio de Fisica y Proteccion Radiologica, Hospital Regional Universitario ' Virgen de las Nieves' , Avda. de las Fuerzas Armadas, 2, E-18014 Granada (Spain)]. E-mail: mvilches@ugr.es; Garcia-Pareja, S. [Servicio de Radiofisica Hospitalaria, Hospital Regional Universitario ' Carlos Haya' , Avda. Carlos Haya, s/n, E-29010 Malaga (Spain)]. E-mail: garciapareja@gmail.com; Guerrero, R. [Servicio de Radiofisica, Hospital Universitario ' San Cecilio' , Avda. Dr. Oloriz, 16, E-18012 Granada (Spain)]. E-mail: rafael.guerrero.alcalde.sspa@juntadeandalucia.es; Anguiano, M. [Departamento de Fisica Atomica, Molecular y Nuclear, Universidad de Granada, E-18071 Granada (Spain)]. E-mail: mangui@ugr.es; Lallena, A.M. [Departamento de Fisica Atomica, Molecular y Nuclear, Universidad de Granada, E-18071 Granada (Spain)]. E-mail: lallena@ugr.es

    2007-01-15

    The Monte Carlo simulation of the electron transport through thin slabs is studied with five general purpose codes: PENELOPE, GEANT3, GEANT4, EGSnrc and MCNPX. The different material foils analyzed in the old experiments of Kulchitsky and Latyshev [L.A. Kulchitsky, G.D. Latyshev, Phys. Rev. 61 (1942) 254] and Hanson et al. [A.O. Hanson, L.H. Lanzl, E.M. Lyman, M.B. Scott, Phys. Rev. 84 (1951) 634] are used to perform the comparison between the Monte Carlo codes. Non-negligible differences are observed in the angular distributions of the transmitted electrons obtained with the some of the codes. The experimental data are reasonably well described by EGSnrc, PENELOPE (v.2005) and GEANT4. A general good agreement is found for EGSnrc and PENELOPE (v.2005) in all the cases analyzed.

  11. A Comparison Between GATE and MCNPX Monte Carlo Codes in Simulation of Medical Linear Accelerator

    Science.gov (United States)

    Sadoughi, Hamid-Reza; Nasseri, Shahrokh; Momennezhad, Mahdi; Sadeghi, Hamid-Reza; Bahreyni-Toosi, Mohammad-Hossein

    2014-01-01

    Radiotherapy dose calculations can be evaluated by Monte Carlo (MC) simulations with acceptable accuracy for dose prediction in complicated treatment plans. In this work, Standard, Livermore and Penelope electromagnetic (EM) physics packages of GEANT4 application for tomographic emission (GATE) 6.1 were compared versus Monte Carlo N-Particle eXtended (MCNPX) 2.6 in simulation of 6 MV photon Linac. To do this, similar geometry was used for the two codes. The reference values of percentage depth dose (PDD) and beam profiles were obtained using a 6 MV Elekta Compact linear accelerator, Scanditronix water phantom and diode detectors. No significant deviations were found in PDD, dose profile, energy spectrum, radial mean energy and photon radial distribution, which were calculated by Standard and Livermore EM models and MCNPX, respectively. Nevertheless, the Penelope model showed an extreme difference. Statistical uncertainty in all the simulations was MCNPX, Standard, Livermore and Penelope models, respectively. Differences between spectra in various regions, in radial mean energy and in photon radial distribution were due to different cross section and stopping power data and not the same simulation of physics processes of MCNPX and three EM models. For example, in the Standard model, the photoelectron direction was sampled from the Gavrila-Sauter distribution, but the photoelectron moved in the same direction of the incident photons in the photoelectric process of Livermore and Penelope models. Using the same primary electron beam, the Standard and Livermore EM models of GATE and MCNPX showed similar output, but re-tuning of primary electron beam is needed for the Penelope model. PMID:24696804

  12. A Particle In Cell code development for high current ion beam transport and plasma simulations

    CERN Document Server

    Joshi, N

    2016-01-01

    A simulation package employing a Particle in Cell (PIC) method is developed to study the high current beam transport and the dynamics of plasmas. This package includes subroutines those are suited for various planned projects at University of Frankfurt. In the framework of the storage ring project (F8SR) the code was written to describe the beam optics in toroidal magnetic fields. It is used to design an injection system for a ring with closed magnetic field lines. The generalized numerical model, in Cartesian coordinates is used to describe the intense ion beam transport through the chopper system in the low energy beam section of the FRANZ project. Especially for the chopper system, the Poisson equation is implemented with irregular geometries. The Particle In Cell model is further upgraded with a Monte Carlo Collision subroutine for simulation of plasma in the volume type ion source.

  13. Full modelling of the MOSAIC animal PET system based on the GATE Monte Carlo simulation code

    Science.gov (United States)

    Merheb, C.; Petegnief, Y.; Talbot, J. N.

    2007-02-01

    Positron emission tomography (PET) systems dedicated to animal imaging are now widely used for biological studies. The scanner performance strongly depends on the design and the characteristics of the system. Many parameters must be optimized like the dimensions and type of crystals, geometry and field-of-view (FOV), sampling, electronics, lightguide, shielding, etc. Monte Carlo modelling is a powerful tool to study the effect of each of these parameters on the basis of realistic simulated data. Performance assessment in terms of spatial resolution, count rates, scatter fraction and sensitivity is an important prerequisite before the model can be used instead of real data for a reliable description of the system response function or for optimization of reconstruction algorithms. The aim of this study is to model the performance of the Philips Mosaic™ animal PET system using a comprehensive PET simulation code in order to understand and describe the origin of important factors that influence image quality. We use GATE, a Monte Carlo simulation toolkit for a realistic description of the ring PET model, the detectors, shielding, cap, electronic processing and dead times. We incorporate new features to adjust signal processing to the Anger logic underlying the Mosaic™ system. Special attention was paid to dead time and energy spectra descriptions. Sorting of simulated events in a list mode format similar to the system outputs was developed to compare experimental and simulated sensitivity and scatter fractions for different energy thresholds using various models of phantoms describing rat and mouse geometries. Count rates were compared for both cylindrical homogeneous phantoms. Simulated spatial resolution was fitted to experimental data for 18F point sources at different locations within the FOV with an analytical blurring function for electronic processing effects. Simulated and measured sensitivities differed by less than 3%, while scatter fractions agreed

  14. Evaluation of atomic electron binding energies for Monte Carlo particle transport

    CERN Document Server

    Pia, Maria Grazia; Batic, Matej; Begalli, Marcia; Kim, Chan Hyeong; Quintieri, Lina; Saracco, Paolo

    2011-01-01

    A survey of atomic binding energies used by general purpose Monte Carlo systems is reported. Various compilations of these parameters have been evaluated; their accuracy is estimated with respect to experimental data. Their effects on physics quantities relevant to Monte Carlo particle transport are highlighted: X-ray fluorescence emission, electron and proton ionization cross sections, and Doppler broadening in Compton scattering. The effects due to different binding energies are quantified with respect to experimental data. The results of the analysis provide quantitative ground for the selection of binding energies to optimize the accuracy of Monte Carlo simulation in experimental use cases. Recommendations on software design dealing with these parameters and on the improvement of data libraries for Monte Carlo simulation are discussed.

  15. NERO - A Post Maximum Supernova Radiation Transport Code

    CERN Document Server

    Maurer, I; Mazzali, P A; Taubenberger, S; Hachinger, S; Kromer, M; Sim, S; Hillebrandt, W

    2011-01-01

    The interpretation of supernova (SN) spectra is essential for deriving SN ejecta properties such as density and composition, which in turn can tell us about their progenitors and the explosion mechanism. A very large number of atomic processes are important for spectrum formation. Several tools for calculating SN spectra exist, but they mainly focus on the very early or late epochs. The intermediate phase, which requires a NLTE treatment of radiation transport has rarely been studied. In this paper we present a new SN radiation transport code, NERO, which can look at those epochs. All the atomic processes are treated in full NLTE, under a steady-state assumption. This is a valid approach between roughly 50 and 500 days after the explosion depending on SN type. This covers the post-maximum photospheric and the early and the intermediate nebular phase. As a test, we compare NERO to the radiation transport code of Jerkstrand et al. (2011) and to the nebular code of Mazzali et al. (2001). All three codes have bee...

  16. NERO- a post-maximum supernova radiation transport code

    Science.gov (United States)

    Maurer, I.; Jerkstrand, A.; Mazzali, P. A.; Taubenberger, S.; Hachinger, S.; Kromer, M.; Sim, S.; Hillebrandt, W.

    2011-12-01

    The interpretation of supernova (SN) spectra is essential for deriving SN ejecta properties such as density and composition, which in turn can tell us about their progenitors and the explosion mechanism. A very large number of atomic processes are important for spectrum formation. Several tools for calculating SN spectra exist, but they mainly focus on the very early or late epochs. The intermediate phase, which requires a non-local thermodynamic equilibrium (NLTE) treatment of radiation transport has rarely been studied. In this paper, we present a new SN radiation transport code, NERO, which can look at those epochs. All the atomic processes are treated in full NLTE, under a steady-state assumption. This is a valid approach between roughly 50 and 500 days after the explosion depending on SN type. This covers the post-maximum photospheric and the early and the intermediate nebular phase. As a test, we compare NERO to the radiation transport code of Jerkstrand, Fransson & Kozma and to the nebular code of Mazzali et al. All three codes have been developed independently and a comparison provides a valuable opportunity to investigate their reliability. Currently, NERO is one-dimensional and can be used for predicting spectra of synthetic explosion models or for deriving SN properties by spectral modelling. To demonstrate this, we study the spectra of the 'normal' Type Ia supernova (SN Ia) 2005cf between 50 and 350 days after the explosion and identify most of the common SN Ia line features at post-maximum epochs.

  17. The Initial Atmospheric Transport (IAT) Code: Description and Validation

    Energy Technology Data Exchange (ETDEWEB)

    Morrow, Charles W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bartel, Timothy James [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-10-01

    The Initial Atmospheric Transport (IAT) computer code was developed at Sandia National Laboratories as part of their nuclear launch accident consequences analysis suite of computer codes. The purpose of IAT is to predict the initial puff/plume rise resulting from either a solid rocket propellant or liquid rocket fuel fire. The code generates initial conditions for subsequent atmospheric transport calculations. The Initial Atmospheric Transfer (IAT) code has been compared to two data sets which are appropriate to the design space of space launch accident analyses. The primary model uncertainties are the entrainment coefficients for the extended Taylor model. The Titan 34D accident (1986) was used to calibrate these entrainment settings for a prototypic liquid propellant accident while the recent Johns Hopkins University Applied Physics Laboratory (JHU/APL, or simply APL) large propellant block tests (2012) were used to calibrate the entrainment settings for prototypic solid propellant accidents. North American Meteorology (NAM )formatted weather data profiles are used by IAT to determine the local buoyancy force balance. The IAT comparisons for the APL solid propellant tests illustrate the sensitivity of the plume elevation to the weather profiles; that is, the weather profile is a dominant factor in determining the plume elevation. The IAT code performed remarkably well and is considered validated for neutral weather conditions.

  18. Thermal-to-fusion neutron convertor and Monte Carlo coupled simulation of deuteron/triton transport and secondary products generation

    Science.gov (United States)

    Wang, Guan-bo; Liu, Han-gang; Wang, Kan; Yang, Xin; Feng, Qi-jie

    2012-09-01

    Thermal-to-fusion neutron convertor has being studied in China Academy of Engineering Physics (CAEP). Current Monte Carlo codes, such as MCNP and GEANT, are inadequate when applied in this multi-step reactions problems. A Monte Carlo tool RSMC (Reaction Sequence Monte Carlo) has been developed to simulate such coupled problem, from neutron absorption, to charged particle ionization and secondary neutron generation. "Forced particle production" variance reduction technique has been implemented to improve the calculation speed distinctly by making deuteron/triton induced secondary product plays a major role. Nuclear data is handled from ENDF or TENDL, and stopping power from SRIM, which described better for low energy deuteron/triton interactions. As a validation, accelerator driven mono-energy 14 MeV fusion neutron source is employed, which has been deeply studied and includes deuteron transport and secondary neutron generation. Various parameters, including fusion neutron angle distribution, average neutron energy at different emission directions, differential and integral energy distributions, are calculated with our tool and traditional deterministic method as references. As a result, we present the calculation results of convertor with RSMC, including conversion ratio of 1 mm 6LiD with a typical thermal neutron (Maxwell spectrum) incidence, and fusion neutron spectrum, which will be used for our experiment.

  19. 3D unstructured-mesh radiation transport codes

    Energy Technology Data Exchange (ETDEWEB)

    Morel, J. [Los Alamos National Lab., NM (United States)

    1997-12-31

    Three unstructured-mesh radiation transport codes are currently being developed at Los Alamos National Laboratory. The first code is ATTILA, which uses an unstructured tetrahedral mesh in conjunction with standard Sn (discrete-ordinates) angular discretization, standard multigroup energy discretization, and linear-discontinuous spatial differencing. ATTILA solves the standard first-order form of the transport equation using source iteration in conjunction with diffusion-synthetic acceleration of the within-group source iterations. DANTE is designed to run primarily on workstations. The second code is DANTE, which uses a hybrid finite-element mesh consisting of arbitrary combinations of hexahedra, wedges, pyramids, and tetrahedra. DANTE solves several second-order self-adjoint forms of the transport equation including the even-parity equation, the odd-parity equation, and a new equation called the self-adjoint angular flux equation. DANTE also offers three angular discretization options: $S{_}n$ (discrete-ordinates), $P{_}n$ (spherical harmonics), and $SP{_}n$ (simplified spherical harmonics). DANTE is designed to run primarily on massively parallel message-passing machines, such as the ASCI-Blue machines at LANL and LLNL. The third code is PERICLES, which uses the same hybrid finite-element mesh as DANTE, but solves the standard first-order form of the transport equation rather than a second-order self-adjoint form. DANTE uses a standard $S{_}n$ discretization in angle in conjunction with trilinear-discontinuous spatial differencing, and diffusion-synthetic acceleration of the within-group source iterations. PERICLES was initially designed to run on workstations, but a version for massively parallel message-passing machines will be built. The three codes will be described in detail and computational results will be presented.

  20. VVER-440 Ex-Core Neutron Transport Calculations by MCNP-5 Code and Comparison with Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Borodkin, Pavel; Khrennikov, Nikolay [Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS) Malaya Krasnoselskaya ul., 2/8, bld. 5, 107140 Moscow (Russian Federation)

    2008-07-01

    Ex-core neutron transport calculations are needed to evaluate radiation loading parameters (neutron fluence, fluence rate and spectra) on the in-vessel equipment, reactor pressure vessel (RPV) and support constructions of VVER type reactors. Due to these parameters are used for reactor equipment life-time assessment, neutron transport calculations should be carried out by precise and reliable calculation methods. In case of RPVs, especially, of first generation VVER-440s, the neutron fluence plays a key role in the prediction of RPV lifetime. Main part of VVER ex-core neutron transport calculations are performed by deterministic and Monte-Carlo methods. This paper deals with precise calculations of the Russian first generation VVER-440 by MCNP-5 code. The purpose of this work was an application of this code for expert calculations, verification of results by comparison with deterministic calculations and validation by neutron activation measured data. Deterministic discrete ordinates DORT code, widely used for RPV neutron dosimetry and many times tested by experiments, was used for comparison analyses. Ex-vessel neutron activation measurements at the VVER-440 NPP have provided space (in azimuth and height directions) and neutron energy (different activation reactions) distributions data for experimental (E) validation of calculated results. Calculational intercomparison (DORT vs. MCNP-5) and comparison with measured values (MCNP-5 and DORT vs. E) have shown agreement within 10-15% for different space points and reaction rates. The paper submits a discussion of results and makes conclusions about practice use of MCNP-5 code for ex-core neutron transport calculations in expert analysis. (authors)

  1. Heavy-ion transport codes for radiotherapy and radioprotection in space

    Energy Technology Data Exchange (ETDEWEB)

    Mancusi, Davide

    2006-06-15

    Simulation of the transport of heavy ions in matter is a field of nuclear science that has recently received attention in view of its importance for some relevant applications. Accelerated heavy ions can, for example, be used to treat cancers (heavy-ion radiotherapy) and show some superior qualities with respect to more conventional treatment systems, like photons (x-rays) or protons. Furthermore, long-term manned space missions (like a possible future mission to Mars) pose the challenge to protect astronauts and equipment on board against the harmful space radiation environment, where heavy ions can be responsible for a significant share of the exposure risk. The high accuracy expected from a transport algorithm (especially in the case of radiotherapy) and the large amount of semi-empirical knowledge necessary to even state the transport problem properly rule out any analytical approach; the alternative is to resort to numerical simulations in order to build treatment-planning systems for cancer or to aid space engineers in shielding design. This thesis is focused on the description of HIBRAC, a one-dimensional deterministic code optimised for radiotherapy, and PHITS (Particle and Heavy- Ion Transport System), a general-purpose three-dimensional Monte-Carlo code. The structure of both codes is outlined and some relevant results are presented. In the case of PHITS, we also report the first results of an ongoing comprehensive benchmarking program for the main components of the code; we present the comparison of partial charge-changing cross sections for a 400 MeV/n {sup 40}Ar beam impinging on carbon, polyethylene, aluminium, copper, tin and lead targets.

  2. Accuracy Evaluation of Oncentra™ TPS in HDR Brachytherapy of Nasopharynx Cancer Using EGSnrc Monte Carlo Code

    Directory of Open Access Journals (Sweden)

    Hadad K

    2015-03-01

    Full Text Available Background: HDR brachytherapy is one of the commonest methods of nasopharyngeal cancer treatment. In this method, depending on how advanced one tumor is, 2 to 6 Gy dose as intracavitary brachytherapy is prescribed. Due to high dose rate and tumor location, accuracy evaluation of treatment planning system (TPS is particularly important. Common methods used in TPS dosimetry are based on computations in a homogeneous phantom. Heterogeneous phantoms, especially patient-specific voxel phantoms can increase dosimetric accuracy. Materials and Methods: In this study, using CT images taken from a patient and ctcreate-which is a part of the DOSXYZnrc computational code, patient-specific phantom was made. Dose distribution was plotted by DOSXYZnrc and compared with TPS one. Also, by extracting the voxels absorbed dose in treatment volume, dosevolume histograms (DVH was plotted and compared with Oncentra™ TPS DVHs. Results: The results from calculations were compared with data from Oncentra™ treatment planning system and it was observed that TPS calculation predicts lower dose in areas near the source, and higher dose in areas far from the source relative to MC code. Absorbed dose values in the voxels also showed that TPS reports D90 value is 40% higher than the Monte Carlo method. Conclusion: Today, most treatment planning systems use TG-43 protocol. This protocol may results in errors such as neglecting tissue heterogeneity, scattered radiation as well as applicator attenuation. Due to these errors, AAPM emphasized departing from TG-43 protocol and approaching new brachytherapy protocol TG-186 in which patient-specific phantom is used and heterogeneities are affected in dosimetry

  3. Using Nuclear Theory, Data and Uncertainties in Monte Carlo Transport Applications

    Energy Technology Data Exchange (ETDEWEB)

    Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-03

    These are slides for a presentation on using nuclear theory, data and uncertainties in Monte Carlo transport applications. The following topics are covered: nuclear data (experimental data versus theoretical models, data evaluation and uncertainty quantification), fission multiplicity models (fixed source applications, criticality calculations), uncertainties and their impact (integral quantities, sensitivity analysis, uncertainty propagation).

  4. Quality control of the treatment planning systems dose calculations in external radiation therapy using the Penelope Monte Carlo code; Controle qualite des systemes de planification dosimetrique des traitements en radiotherapie externe au moyen du code Monte-Carlo Penelope

    Energy Technology Data Exchange (ETDEWEB)

    Blazy-Aubignac, L

    2007-09-15

    The treatment planning systems (T.P.S.) occupy a key position in the radiotherapy service: they realize the projected calculation of the dose distribution and the treatment duration. Traditionally, the quality control of the calculated distribution doses relies on their comparisons with dose distributions measured under the device of treatment. This thesis proposes to substitute these dosimetry measures to the profile of reference dosimetry calculations got by the Penelope Monte-Carlo code. The Monte-Carlo simulations give a broad choice of test configurations and allow to envisage a quality control of dosimetry aspects of T.P.S. without monopolizing the treatment devices. This quality control, based on the Monte-Carlo simulations has been tested on a clinical T.P.S. and has allowed to simplify the quality procedures of the T.P.S.. This quality control, in depth, more precise and simpler to implement could be generalized to every center of radiotherapy. (N.C.)

  5. HERMES: a Monte Carlo Code for the Propagation of Ultra-High Energy Nuclei

    CERN Document Server

    De Domenico, Manlio; Settimo, Mariangela

    2013-01-01

    Although the recent experimental efforts to improve the observation of Ultra-High Energy Cosmic Rays (UHECRs) above $10^{18}$ eV, the origin and the composition of such particles is still unknown. In this work, we present the novel Monte Carlo code (HERMES) simulating the propagation of UHE nuclei, in the energy range between $10^{16}$ and $10^{22}$ eV, accounting for propagation in the intervening extragalactic and Galactic magnetic fields and nuclear interactions with relic photons of the extragalactic background radiation. In order to show the potential applications of HERMES for astroparticle studies, we estimate the expected flux of UHE nuclei in different astrophysical scenarios, the GZK horizons and we show the expected arrival direction distributions in the presence of turbulent extragalactic magnetic fields. A stable version of HERMES will be released in the next future for public use together with libraries of already propagated nuclei to allow the community to perform mass composition and energy sp...

  6. Design and Simulation of Photoneutron Source by MCNPX Monte Carlo Code for Boron Neutron Capture Therapy

    Directory of Open Access Journals (Sweden)

    Mona Zolfaghari

    2015-07-01

    Full Text Available Introduction Electron linear accelerator (LINAC can be used for neutron production in Boron Neutron Capture Therapy (BNCT. BNCT is an external radiotherapeutic method for the treatment of some cancers. In this study, Varian 2300 C/D LINAC was simulated as an electron accelerator-based photoneutron source to provide a suitable neutron flux for BNCT. Materials and Methods Photoneutron sources were simulated, using MCNPX Monte Carlo code. In this study, a 20 MeV LINAC was utilized for electron-photon reactions. After the evaluation of cross-sections and threshold energies, lead (Pb, uranium (U and beryllium deuteride (BeD2were selected as photoneutron sources. Results According to the simulation results, optimized photoneutron sources with a compact volume and photoneutron yields of 107, 108 and 109 (n.cm-2.s-1 were obtained for Pb, U and BeD2 composites. Also, photoneutrons increased by using enriched U (10-60% as an electron accelerator-based photoneutron source. Conclusion Optimized photoneutron sources were obtained with compact sizes of 107, 108 and 109 (n.cm-2.s-1, respectively. These fluxs can be applied for BNCT by decelerating fast neutrons and using a suitable beam-shaping assembly, surrounding electron-photon and photoneutron sources.

  7. Thyroid cell irradiation by radioiodines: a new Monte Carlo electron track-structure code

    Energy Technology Data Exchange (ETDEWEB)

    Champion, Christophe [Universite Paul Verlaine-Metz (France). Lab. de Physique Moleculaire et des Collisions]. E-mail: champion@univ-metz.fr; Elbast, Mouhamad; Colas-Linhart, Nicole [Universite Paris 7 (France). Faculte de Medecine. Lab. de Biophysique; Ting-Di Wu [INSERM U759, Orsay (France). Institut Curie Recherche. Imagerie Integrative

    2007-09-15

    The most significant impact of the Chernobyl accident is the increased incidence of thyroid cancer among children who were exposed to short-lived radioiodines and 131-iodine. In order to accurately estimate the radiation dose provided by these radioiodines, it is necessary to know where iodine is incorporated. To do that, the distribution at the cellular level of newly organified iodine in the immature rat thyroid was performed using secondary ion mass microscopy (NanoSIMS{sup 50}). Actual dosimetric models take only into account the averaged energy and range of beta particles of the radio-elements and may, therefore, imperfectly describe the real distribution of dose deposit at the microscopic level around the point sources. Our approach is radically different since based on a track-structure Monte Carlo code allowing following-up of electrons down to low energies ({approx}= 10 eV) what permits a nanometric description of the irradiation physics. The numerical simulations were then performed by modelling the complete disintegrations of the short-lived iodine isotopes as well as of {sup 131}I in new born rat thyroids in order to take into account accurate histological and biological data for the thyroid gland. (author)

  8. Deep-penetration calculation for the ISIS target station shielding using the MARS Monte Carlo code

    CERN Document Server

    Nunomiya, T; Nakamura, T; Nakao, N

    2002-01-01

    A calculation of neutron penetration through a thick shield was performed with a three-dimensional multi-layer technique using the MARS14(02) Monte Carlo code to compare with the experimental shielding data in 1998 at the ISIS spallation neutron source facility. In this calculation, secondary particles from a tantalum target bombarded by 800-MeV protons were transmitted through a bulk shield of approximately 3-m-thick iron and 1-m-thick concrete. To accomplish this deep-penetration calculation with good statistics, the following three techniques were used in this study. First, the geometry of the bulk shield was three-dimensionally divided into several layers of about 50-cm thickness, and a step-by-step calculation was carried out to multiply the number of penetrated particles at the boundaries between the layers. Second, the source particles in the layers were divided into two parts to maintain the statistical balance on the spatial-flux distribution. Third, only high-energy particles above 20 MeV were trans...

  9. Particle, momentum and thermal transport in the PTRANSP code

    Science.gov (United States)

    Bateman, G.; Halpern, F. D.; Kritz, A. H.; Pankin, A. Y.; Rafiq, T.; McCune, D. C.; Budny, R. V.; Indireshkumar, K.

    2008-11-01

    The combined effects of particle, momentum and thermal transport are investigated in tokamak discharges using a coupled system of transport equations implemented in the PTRANSP integrated modeling code. The magnetic diffusion equation is advanced separately, along with the evolution of the equilibrium. Simulations are carried out using theory-based models to compute transport, sources and sinks. Boundary conditions are either read from data or computed using a pedestal model for H-mode discharges. Different techniques are explored for controlling numerical problems [1] in time-dependent simulations that include sawtooth oscillations and other rapid changes in the profiles. Results for the density, temperature and toroidal angular velocity profiles are compared with experimental data. [1] S.C. Jardin et al, ``On 1D diffusion problems with a gradient-dependent diffusion coefficient''; G.V. Pereverzev and G. Corrigan, ``Stable numeric scheme for diffusion equation with a stiff transport''; both papers to appear in Comp. Phys. Comm. (2008).

  10. Monte Carlo path sampling approach to modeling aeolian sediment transport

    Science.gov (United States)

    Hardin, E. J.; Mitasova, H.; Mitas, L.

    2011-12-01

    Coastal communities and vital infrastructure are subject to coastal hazards including storm surge and hurricanes. Coastal dunes offer protection by acting as natural barriers from waves and storm surge. During storms, these landforms and their protective function can erode; however, they can also erode even in the absence of storms due to daily wind and waves. Costly and often controversial beach nourishment and coastal construction projects are common erosion mitigation practices. With a more complete understanding of coastal morphology, the efficacy and consequences of anthropogenic activities could be better predicted. Currently, the research on coastal landscape evolution is focused on waves and storm surge, while only limited effort is devoted to understanding aeolian forces. Aeolian transport occurs when the wind supplies a shear stress that exceeds a critical value, consequently ejecting sand grains into the air. If the grains are too heavy to be suspended, they fall back to the grain bed where the collision ejects more grains. This is called saltation and is the salient process by which sand mass is transported. The shear stress required to dislodge grains is related to turbulent air speed. Subsequently, as sand mass is injected into the air, the wind loses speed along with its ability to eject more grains. In this way, the flux of saltating grains is itself influenced by the flux of saltating grains and aeolian transport becomes nonlinear. Aeolian sediment transport is difficult to study experimentally for reasons arising from the orders of magnitude difference between grain size and dune size. It is difficult to study theoretically because aeolian transport is highly nonlinear especially over complex landscapes. Current computational approaches have limitations as well; single grain models are mathematically simple but are computationally intractable even with modern computing power whereas cellular automota-based approaches are computationally efficient

  11. New Parallel computing framework for radiation transport codes

    Energy Technology Data Exchange (ETDEWEB)

    Kostin, M.A.; /Michigan State U., NSCL; Mokhov, N.V.; /Fermilab; Niita, K.; /JAERI, Tokai

    2010-09-01

    A new parallel computing framework has been developed to use with general-purpose radiation transport codes. The framework was implemented as a C++ module that uses MPI for message passing. The module is significantly independent of radiation transport codes it can be used with, and is connected to the codes by means of a number of interface functions. The framework was integrated with the MARS15 code, and an effort is under way to deploy it in PHITS. Besides the parallel computing functionality, the framework offers a checkpoint facility that allows restarting calculations with a saved checkpoint file. The checkpoint facility can be used in single process calculations as well as in the parallel regime. Several checkpoint files can be merged into one thus combining results of several calculations. The framework also corrects some of the known problems with the scheduling and load balancing found in the original implementations of the parallel computing functionality in MARS15 and PHITS. The framework can be used efficiently on homogeneous systems and networks of workstations, where the interference from the other users is possible.

  12. New Parallel computing framework for radiation transport codes

    CERN Document Server

    Kostin, M A; Niita, K

    2012-01-01

    A new parallel computing framework has been developed to use with general-purpose radiation transport codes. The framework was implemented as a C++ module that uses MPI for message passing. The module is significantly independent of radiation transport codes it can be used with, and is connected to the codes by means of a number of interface functions. The frame work was integrated with the MARS15 code, and an effort is under way to deploy it in PHITS. Besides the parallel computing functionality, the framework offers a checkpoint facility that allows restarting calculations with a saved checkpoint file. The checkpoint facility can be used in single process calculations as well as in the parallel regime. Several checkpoint files can be merged into one thus combining results of several calculations. The framework also corrects some of the known problems with the sch eduling and load balancing found in the original implementations of the parallel computing functionality in MARS15 and PHITS. The framework can be...

  13. The specific purpose Monte Carlo code McENL for simulating the response of epithermal neutron lifetime well logging tools

    Energy Technology Data Exchange (ETDEWEB)

    Prettyman, T.H.; Gardner, R.P.; Verghese, K. (North Carolina State Univ., Raleigh, NC (United States). Center for Engineering Applications and Radioisotopes)

    1993-08-01

    A new specific purpose Monte Carlo code called McENL for modeling the time response of epithermal neutron lifetime tools is described. The code was developed so that the Monte Carlo neophyte can easily use it. A minimum amount of input preparation is required and specified fixed values of the parameters used to control the code operation can be used. The weight windows technique, employing splitting and Russian Roulette, is used with an automated importance function based on the solution of an adjoint diffusion model to improve the code efficiency. Complete composition and density correlated sampling is also included in the code and can be used to study the effect on tool response of small variations in the formation, borehole, or logging tool composition and density. An illustration of the latter application is given here for the density of a thermal neutron filter. McENL was benchmarked against test-pit data for the Mobil pulsed neutron porosity (PNP) tool and found to be very accurate. Results of the experimental validation and details of code performance are presented.

  14. High energy particle transport code NMTC/JAM

    Energy Technology Data Exchange (ETDEWEB)

    Niita, Koji [Research Organization for Information Science and Technology, Tokai, Ibaraki (Japan); Meigo, Shin-ichiro; Takada, Hiroshi; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    We have developed a high energy particle transport code NMTC/JAM, which is an upgraded version of NMTC/JAERI97. The applicable energy range of NMTC/JAM is extended in principle up to 200 GeV for nucleons and mesons by introducing the high energy nuclear reaction code JAM for the intra-nuclear cascade part. For the evaporation and fission process, we have also implemented a new model, GEM, by which the light nucleus production from the excited residual nucleus can be described. According to the extension of the applicable energy, we have upgraded the nucleon-nucleus non-elastic, elastic and differential elastic cross section data by employing new systematics. In addition, the particle transport in a magnetic field has been implemented for the beam transport calculations. In this upgrade, some new tally functions are added and the format of input of data has been improved very much in a user friendly manner. Due to the implementation of these new calculation functions and utilities, consequently, NMTC/JAM enables us to carry out reliable neutronics study of a large scale target system with complex geometry more accurately and easily than before. This report serves as a user manual of the code. (author)

  15. Monte Carlo transport simulation of velocity undershoot in zinc blende and wurtzite InN

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shulong; Liu, Hongxia; Gao, Bo; Zhuo, Qingqing [School of Microelectronics, Key Laboratory of Wide Band-gap Semiconductor Materials and Device, Xidian University, Xi& #x27; an, 710071 (China)

    2012-09-15

    Velocity undershoot in zinc blende (ZB) and wurtzite (WZ) InN is investigated by ensemble Monte Carlo (EMC) calculation. The results show that velocity undershoot arises from the relatively long energy relaxation time compared with momentum. Monte Carlo transport simulations over wide range of electric fields is presented in the paper. The results show that velocity undershoot impacts the electron transport greatly, compared with velocity overshoot, when the electric field changes quickly with time and space. A comparison study between WZ and ZB InN shows that WZ InN has more advantages in device applications due to its excellent electron transport properties. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  16. Performance Analysis of Korean Liquid metal type TBM based on Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C. H.; Han, B. S.; Park, H. J.; Park, D. K. [Seoul National Univ., Seoul (Korea, Republic of)

    2007-01-15

    The objective of this project is to analyze a nuclear performance of the Korean HCML(Helium Cooled Molten Lithium) TBM(Test Blanket Module) which will be installed in ITER(International Thermonuclear Experimental Reactor). This project is intended to analyze a neutronic design and nuclear performances of the Korean HCML ITER TBM through the transport calculation of MCCARD. In detail, we will conduct numerical experiments for analyzing the neutronic design of the Korean HCML TBM and the DEMO fusion blanket, and improving the nuclear performances. The results of the numerical experiments performed in this project will be utilized further for a design optimization of the Korean HCML TBM. In this project, Monte Carlo transport calculations for evaluating TBR (Tritium Breeding Ratio) and EMF (Energy Multiplication factor) were conducted to analyze a nuclear performance of the Korean HCML TBM. The activation characteristics and shielding performances for the Korean HCML TBM were analyzed using ORIGEN and MCCARD. We proposed the neutronic methodologies for analyzing the nuclear characteristics of the fusion blanket, which was applied to the blanket analysis of a DEMO fusion reactor. In the results, the TBR of the Korean HCML ITER TBM is 0.1352 and the EMF is 1.362. Taking into account a limitation for the Li amount in ITER TBM, it is expected that tritium self-sufficiency condition can be satisfied through a change of the Li quantity and enrichment. In the results of activation and shielding analysis, the activity drops to 1.5% of the initial value and the decay heat drops to 0.02% of the initial amount after 10 years from plasma shutdown.

  17. Comparative Dosimetric Estimates of a 25 keV Electron Micro-beam with three Monte Carlo Codes

    Energy Technology Data Exchange (ETDEWEB)

    Mainardi, Enrico; Donahue, Richard J.; Blakely, Eleanor A.

    2002-09-11

    The calculations presented compare the different performances of the three Monte Carlo codes PENELOPE-1999, MCNP-4C and PITS, for the evaluation of Dose profiles from a 25 keV electron micro-beam traversing individual cells. The overall model of a cell is a water cylinder equivalent for the three codes but with a different internal scoring geometry: hollow cylinders for PENELOPE and MCNP, whereas spheres are used for the PITS code. A cylindrical cell geometry with scoring volumes with the shape of hollow cylinders was initially selected for PENELOPE and MCNP because of its superior simulation of the actual shape and dimensions of a cell and for its improved computer-time efficiency if compared to spherical internal volumes. Some of the transfer points and energy transfer that constitute a radiation track may actually fall in the space between spheres, that would be outside the spherical scoring volume. This internal geometry, along with the PENELOPE algorithm, drastically reduced the computer time when using this code if comparing with event-by-event Monte Carlo codes like PITS. This preliminary work has been important to address dosimetric estimates at low electron energies. It demonstrates that codes like PENELOPE can be used for Dose evaluation, even with such small geometries and energies involved, which are far below the normal use for which the code was created. Further work (initiated in Summer 2002) is still needed however, to create a user-code for PENELOPE that allows uniform comparison of exact cell geometries, integral volumes and also microdosimetric scoring quantities, a field where track-structure codes like PITS, written for this purpose, are believed to be superior.

  18. Analysis of the dead layer of a detector of germanium with code ultrapure Monte Carlo SWORD-GEANT; Analisis del dead layer de un detector de germanio ultrapuro con el codigo de Monte Carlo SWORDS-GEANT

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, S.; Querol, A.; Ortiz, J.; Rodenas, J.; Verdu, G.

    2014-07-01

    In this paper the use of Monte Carlo code SWORD-GEANT is proposed to simulate an ultra pure germanium detector High Purity Germanium detector (HPGe) detector ORTEC specifically GMX40P4, coaxial geometry. (Author)

  19. The Monte Carlo approach to transport modeling in deca-nanometer MOSFETs

    Science.gov (United States)

    Sangiorgi, Enrico; Palestri, Pierpaolo; Esseni, David; Fiegna, Claudio; Selmi, Luca

    2008-09-01

    In this paper, we review recent developments of the Monte Carlo approach to the simulation of semi-classical carrier transport in nano-MOSFETs, with particular focus on the inclusion of quantum-mechanical effects in the simulation (using either the multi-subband approach or quantum corrections to the electrostatic potential) and on the numerical stability issues related to the coupling of the transport with the Poisson equation. Selected applications are presented, including the analysis of quasi-ballistic transport, the determination of the RF characteristics of deca-nanometric MOSFETs, and the study of non-conventional device structures and channel materials.

  20. Verification of ARES transport code system with TAKEDA benchmarks

    Science.gov (United States)

    Zhang, Liang; Zhang, Bin; Zhang, Penghe; Chen, Mengteng; Zhao, Jingchang; Zhang, Shun; Chen, Yixue

    2015-10-01

    Neutron transport modeling and simulation are central to many areas of nuclear technology, including reactor core analysis, radiation shielding and radiation detection. In this paper the series of TAKEDA benchmarks are modeled to verify the critical calculation capability of ARES, a discrete ordinates neutral particle transport code system. SALOME platform is coupled with ARES to provide geometry modeling and mesh generation function. The Koch-Baker-Alcouffe parallel sweep algorithm is applied to accelerate the traditional transport calculation process. The results show that the eigenvalues calculated by ARES are in excellent agreement with the reference values presented in NEACRP-L-330, with a difference less than 30 pcm except for the first case of model 3. Additionally, ARES provides accurate fluxes distribution compared to reference values, with a deviation less than 2% for region-averaged fluxes in all cases. All of these confirms the feasibility of ARES-SALOME coupling and demonstrate that ARES has a good performance in critical calculation.

  1. EleCa: A Monte Carlo code for the propagation of extragalactic photons at ultra-high energy

    Energy Technology Data Exchange (ETDEWEB)

    Settimo, Mariangela [University of Siegen (Germany); De Domenico, Manlio [Laboratory of Complex Systems, Scuola Superiore di Catania and INFN (Italy); Lyberis, Haris [Federal University of Rio de Janeiro (Brazil)

    2013-06-15

    Ultra high energy photons, above 10{sup 17}–10{sup 18}eV, can interact with the extragalactic background radiation leading to the development of electromagnetic cascades. A Monte Carlo code to simulate the electromagnetic cascades initiated by high-energy photons and electrons is presented. Results from simulations and their impact on the predicted flux at Earth are discussed in different astrophysical scenarios.

  2. Coupling MCNP-DSP and LAHET Monte Carlo codes for designing subcriticality monitors for accelerator-driven systems

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.; Perez, R. [Oak Ridge National Lab., TN (United States); Rugama, Y.; Munoz-Cobo, J.L. [Poly. Tech. Univ. of Valencia (Spain). Chemical and Nuclear Engineering Dept.

    2001-07-01

    The design of reactivity monitoring systems for accelerator-driven systems must be investigated to ensure that such systems remain subcritical during operation. The Monte Carlo codes LAHET and MCNP-DSP were combined together to facilitate the design of reactivity monitoring systems. The coupling of LAHET and MCNP-DSP provides a tool that can be used to simulate a variety of subcritical measurements such as the pulsed neutron, Rossi-{alpha}, or noise analysis measurements. (orig.)

  3. Coupling MCNP-DSP and LAHET Monte Carlo Codes for Designing Subcriticality Monitors for Accelerator-Driven Systems

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.; Rugama, Y. Munoz-Cobos, J.; Perez, R.

    2000-10-23

    The design of reactivity monitoring systems for accelerator-driven systems must be investigated to ensure that such systems remain subcritical during operation. The Monte Carlo codes LAHET and MCNP-DSP were combined together to facilitate the design of reactivity monitoring systems. The coupling of LAHET and MCNP-DSP provides a tool that can be used to simulate a variety of subcritical measurements such as the pulsed neutron, Rossi-{alpha}, or noise analysis measurements.

  4. Optimization of GATE and PHITS Monte Carlo code parameters for spot scanning proton beam based on simulation with FLUKA general-purpose code

    Energy Technology Data Exchange (ETDEWEB)

    Kurosu, Keita [Department of Radiation Oncology, Indiana University School of Medicine, Indianapolis, IN 46202 (United States); Department of Radiation Oncology, Osaka University Graduate School of Medicine, Suita, Osaka 565-0871 (Japan); Department of Radiology, Osaka University Hospital, Suita, Osaka 565-0871 (Japan); Das, Indra J. [Department of Radiation Oncology, Indiana University School of Medicine, Indianapolis, IN 46202 (United States); Moskvin, Vadim P. [Department of Radiation Oncology, Indiana University School of Medicine, Indianapolis, IN 46202 (United States); Department of Radiation Oncology, St. Jude Children’s Research Hospital, Memphis, TN 38105 (United States)

    2016-01-15

    Spot scanning, owing to its superior dose-shaping capability, provides unsurpassed dose conformity, in particular for complex targets. However, the robustness of the delivered dose distribution and prescription has to be verified. Monte Carlo (MC) simulation has the potential to generate significant advantages for high-precise particle therapy, especially for medium containing inhomogeneities. However, the inherent choice of computational parameters in MC simulation codes of GATE, PHITS and FLUKA that is observed for uniform scanning proton beam needs to be evaluated. This means that the relationship between the effect of input parameters and the calculation results should be carefully scrutinized. The objective of this study was, therefore, to determine the optimal parameters for the spot scanning proton beam for both GATE and PHITS codes by using data from FLUKA simulation as a reference. The proton beam scanning system of the Indiana University Health Proton Therapy Center was modeled in FLUKA, and the geometry was subsequently and identically transferred to GATE and PHITS. Although the beam transport is managed by spot scanning system, the spot location is always set at the center of a water phantom of 600 × 600 × 300 mm{sup 3}, which is placed after the treatment nozzle. The percentage depth dose (PDD) is computed along the central axis using 0.5 × 0.5 × 0.5 mm{sup 3} voxels in the water phantom. The PDDs and the proton ranges obtained with several computational parameters are then compared to those of FLUKA, and optimal parameters are determined from the accuracy of the proton range, suppressed dose deviation, and computational time minimization. Our results indicate that the optimized parameters are different from those for uniform scanning, suggesting that the gold standard for setting computational parameters for any proton therapy application cannot be determined consistently since the impact of setting parameters depends on the proton irradiation

  5. Uncertainty analysis in the simulation of an HPGe detector using the Monte Carlo Code MCNP5

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, Sergio; Pozuelo, Fausto; Querol, Andrea; Verdu, Gumersindo; Rodenas, Jose, E-mail: sergalbe@upv.es [Universitat Politecnica de Valencia, Valencia, (Spain). Instituto de Seguridad Industrial, Radiofisica y Medioambiental (ISIRYM); Ortiz, J. [Universitat Politecnica de Valencia, Valencia, (Spain). Servicio de Radiaciones. Lab. de Radiactividad Ambiental; Pereira, Claubia [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2013-07-01

    A gamma spectrometer including an HPGe detector is commonly used for environmental radioactivity measurements. Many works have been focused on the simulation of the HPGe detector using Monte Carlo codes such as MCNP5. However, the simulation of this kind of detectors presents important difficulties due to the lack of information from manufacturers and due to loss of intrinsic properties in aging detectors. Some parameters such as the active volume or the Ge dead layer thickness are many times unknown and are estimated during simulations. In this work, a detailed model of an HPGe detector and a petri dish containing a certified gamma source has been done. The certified gamma source contains nuclides to cover the energy range between 50 and 1800 keV. As a result of the simulation, the Pulse Height Distribution (PHD) is obtained and the efficiency curve can be calculated from net peak areas and taking into account the certified activity of the source. In order to avoid errors due to the net area calculation, the simulated PHD is treated using the GammaVision software. On the other hand, it is proposed to use the Noether-Wilks formula to do an uncertainty analysis of model with the main goal of determining the efficiency curve of this detector and its associated uncertainty. The uncertainty analysis has been focused on dead layer thickness at different positions of the crystal. Results confirm the important role of the dead layer thickness in the low energy range of the efficiency curve. In the high energy range (from 300 to 1800 keV) the main contribution to the absolute uncertainty is due to variations in the active volume. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Panettieri, Vanessa [Institut de Tecniques Energetiques, Universitat Politecnica de Catalunya, Diagonal 647, 08028 Barcelona (Spain); Duch, Maria Amor [Institut de Tecniques Energetiques, Universitat Politecnica de Catalunya, Diagonal 647, 08028 Barcelona (Spain); Jornet, Nuria [Servei de RadiofIsica i Radioproteccio, Hospital de la Santa Creu i San Pau Sant Antoni Maria Claret 167, 08025 Barcelona (Spain); Ginjaume, Merce [Institut de Tecniques Energetiques, Universitat Politecnica de Catalunya, Diagonal 647, 08028 Barcelona (Spain); Carrasco, Pablo [Servei de RadiofIsica i Radioproteccio, Hospital de la Santa Creu i San Pau Sant Antoni Maria Claret 167, 08025 Barcelona (Spain); Badal, Andreu [Institut de Tecniques Energetiques, Universitat Politecnica de Catalunya, Diagonal 647, 08028 Barcelona (Spain); Ortega, Xavier [Institut de Tecniques Energetiques, Universitat Politecnica de Catalunya, Diagonal 647, 08028 Barcelona (Spain); Ribas, Montserrat [Servei de RadiofIsica i Radioproteccio, Hospital de la Santa Creu i San Pau Sant Antoni Maria Claret 167, 08025 Barcelona (Spain)

    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 and Nielsen TN-502-RD MOSFETs have a very small sensitive area of 0.04 mm{sup 2} and a thickness of 0.5 {mu}m 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(TM) 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(TM) 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

  7. Observing gas and dust in simulations of star formation with Monte Carlo radiation transport on Voronoi meshes

    CERN Document Server

    Hubber, D A; Dale, J

    2015-01-01

    Ionising feedback from massive stars dramatically affects the interstellar medium local to star forming regions. Numerical simulations are now starting to include enough complexity to produce morphologies and gas properties that are not too dissimilar from observations. The comparison between the density fields produced by hydrodynamical simulations and observations at given wavelengths relies however on photoionisation/chemistry and radiative transfer calculations. We present here an implementation of Monte Carlo radiation transport through a Voronoi tessellation in the photoionisation and dust radiative transfer code MOCASSIN. We show for the first time a synthetic spectrum and synthetic emission line maps of an hydrodynamical simulation of a molecular cloud affected by massive stellar feedback. We show that the approach on which previous work is based, which remapped hydrodynamical density fields onto Cartesian grids before performing radiative transfer/photoionisation calculations, results in significant ...

  8. Validation of a personalized dosimetric evaluation tool (Oedipe) for targeted radiotherapy based on the Monte Carlo MCNPX code.

    Science.gov (United States)

    Chiavassa, S; Aubineau-Lanièce, I; Bitar, A; Lisbona, A; Barbet, J; Franck, D; Jourdain, J R; Bardiès, M

    2006-02-07

    Dosimetric studies are necessary for all patients treated with targeted radiotherapy. In order to attain the precision required, we have developed Oedipe, a dosimetric tool based on the MCNPX Monte Carlo code. The anatomy of each patient is considered in the form of a voxel-based geometry created using computed tomography (CT) images or magnetic resonance imaging (MRI). Oedipe enables dosimetry studies to be carried out at the voxel scale. Validation of the results obtained by comparison with existing methods is complex because there are multiple sources of variation: calculation methods (different Monte Carlo codes, point kernel), patient representations (model or specific) and geometry definitions (mathematical or voxel-based). In this paper, we validate Oedipe by taking each of these parameters into account independently. Monte Carlo methodology requires long calculation times, particularly in the case of voxel-based geometries, and this is one of the limits of personalized dosimetric methods. However, our results show that the use of voxel-based geometry as opposed to a mathematically defined geometry decreases the calculation time two-fold, due to an optimization of the MCNPX2.5e code. It is therefore possible to envisage the use of Oedipe for personalized dosimetry in the clinical context of targeted radiotherapy.

  9. Applying Advanced Neutron Transport Calculations for Improving Fuel Performance Codes

    Energy Technology Data Exchange (ETDEWEB)

    Botazzoli, P.; Luzzi, L. [Politecnico di Milano, Department of Energy, Nuclear Engineering Division - CeSNEF, Milano (Italy); Schubert, A.; Van Uffelen, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Karlsruhe (Germany); Haeck, W. [Institute de Radioprotection et de Surete Nucleaire, Fontenay-aux-Roses (France)

    2009-06-15

    TRANSURANUS is a computer code for the thermal and mechanical analysis of fuel rods in nuclear reactors. As part of the code, the TUBRNP model calculates the local concentration of the actinides (U, Pu, Am, Cm), the main fission products (Xe, Kr, Cs and Nd) and {sup 4}He produced during the irradiation as a function of the radial position across a fuel pellet (radial profiles). These local quantities are required for the determination of the local power density, the local burn-up, and the source term of fission products and other inert gases. In previous works the neutronic code ALEPH has been used to validate the models for the actinides and fission products concentrations in UO{sub 2} fuels. A similar approach has been adopted in the present work for verifying the Helium production. The present paper focuses on the modelling of the Helium production in PWR oxide fuels (MOX and UO{sub 2}). A reliable prediction of the Helium production and release in LWR oxide fuels is of great interest in case of increasing burn-up, linear heat generation rates and Plutonium content. The contribution of the Helium released plays a fundamental role in the gap pressure and subsequently in the mechanical behaviour of the fuel rod, in particular during the storage of the high burn-up spent fuel. Helium is produced in oxide fuels by three main paths: (i) alpha decay of the actinides (the main contribution is due to {sup 242}Cm, {sup 238}Pu and {sup 244}Cm); (ii) (n,{alpha}) reactions; and (iii) ternary fission. In the present work, the contributions due to ternary fission and the (n,{alpha}) reaction on {sup 16}O as well as some refinements in the {sup 241}Am burn-up chain have been included in TUBRNP. The VESTA neutronic code has been used for the validation of the He production model. The generic VESTA Monte Carlo depletion interface developed at IRSN allows us to couple different Monte Carlo codes with a depletion module. It currently allows for combining the ORIGEN 2.2 isotope

  10. Application of Photon Transport Monte Carlo Module with GPU-based Parallel System

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Je [Sejong University, Seoul (Korea, Republic of); Shon, Heejeong [Golden Eng. Co. LTD, Seoul (Korea, Republic of); Lee, Donghak [CoCo Link Inc., Seoul (Korea, Republic of)

    2015-05-15

    In general, it takes lots of computing time to get reliable results in Monte Carlo simulations especially in deep penetration problems with a thick shielding medium. To mitigate such a weakness of Monte Carlo methods, lots of variance reduction algorithms are proposed including geometry splitting and Russian roulette, weight windows, exponential transform, and forced collision, etc. Simultaneously, advanced computing hardware systems such as GPU(Graphics Processing Units)-based parallel machines are used to get a better performance of the Monte Carlo simulation. The GPU is much easier to access and to manage when comparing a CPU cluster system. It also becomes less expensive these days due to enhanced computer technology. There, lots of engineering areas adapt GPU-bases massive parallel computation technique. based photon transport Monte Carlo method. It provides almost 30 times speedup without any optimization and it is expected almost 200 times with fully supported GPU system. It is expected that GPU system with advanced parallelization algorithm will contribute successfully for development of the Monte Carlo module which requires quick and accurate simulations.

  11. COOL: A code for Dynamic Monte Carlo Simulation of molecular dynamics

    Science.gov (United States)

    Barletta, Paolo

    2012-02-01

    Cool is a program to simulate evaporative and sympathetic cooling for a mixture of two gases co-trapped in an harmonic potential. The collisions involved are assumed to be exclusively elastic, and losses are due to evaporation from the trap. Each particle is followed individually in its trajectory, consequently properties such as spatial densities or energy distributions can be readily evaluated. The code can be used sequentially, by employing one output as input for another run. The code can be easily generalised to describe more complicated processes, such as the inclusion of inelastic collisions, or the possible presence of more than two species in the trap. New version program summaryProgram title: COOL Catalogue identifier: AEHJ_v2_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEHJ_v2_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 1 097 733 No. of bytes in distributed program, including test data, etc.: 18 425 722 Distribution format: tar.gz Programming language: C++ Computer: Desktop Operating system: Linux RAM: 500 Mbytes Classification: 16.7, 23 Catalogue identifier of previous version: AEHJ_v1_0 Journal reference of previous version: Comput. Phys. Comm. 182 (2011) 388 Does the new version supersede the previous version?: Yes Nature of problem: Simulation of the sympathetic process occurring for two molecular gases co-trapped in a deep optical trap. Solution method: The Direct Simulation Monte Carlo method exploits the decoupling, over a short time period, of the inter-particle interaction from the trapping potential. The particle dynamics is thus exclusively driven by the external optical field. The rare inter-particle collisions are considered with an acceptance/rejection mechanism, that is, by comparing a random number to the collisional probability

  12. An OpenCL-based Monte Carlo dose calculation engine (oclMC) for coupled photon-electron transport

    CERN Document Server

    Tian, Zhen; Folkerts, Michael; Qin, Nan; Jiang, Steve B; Jia, Xun

    2015-01-01

    Monte Carlo (MC) method has been recognized the most accurate dose calculation method for radiotherapy. However, its extremely long computation time impedes clinical applications. Recently, a lot of efforts have been made to realize fast MC dose calculation on GPUs. Nonetheless, most of the GPU-based MC dose engines were developed in NVidia CUDA environment. This limits the code portability to other platforms, hindering the introduction of GPU-based MC simulations to clinical practice. The objective of this paper is to develop a fast cross-platform MC dose engine oclMC using OpenCL environment for external beam photon and electron radiotherapy in MeV energy range. Coupled photon-electron MC simulation was implemented with analogue simulations for photon transports and a Class II condensed history scheme for electron transports. To test the accuracy and efficiency of our dose engine oclMC, we compared dose calculation results of oclMC and gDPM, our previously developed GPU-based MC code, for a 15 MeV electron ...

  13. On the Way to Future's High Energy Particle Physics Transport Code

    CERN Document Server

    Bíró, Gábor; Futó, Endre

    2015-01-01

    High Energy Physics (HEP) needs a huge amount of computing resources. In addition data acquisition, transfer, and analysis require a well developed infrastructure too. In order to prove new physics disciplines it is required to higher the luminosity of the accelerator facilities, which produce more-and-more data in the experimental detectors. Both testing new theories and detector R&D are based on complex simulations. Today have already reach that level, the Monte Carlo detector simulation takes much more time than real data collection. This is why speed up of the calculations and simulations became important in the HEP community. The Geant Vector Prototype (GeantV) project aims to optimize the most-used particle transport code applying parallel computing and to exploit the capabilities of the modern CPU and GPU architectures as well. With the maximized concurrency at multiple levels the GeantV is intended to be the successor of the Geant4 particle transport code that has been used since two decades succe...

  14. Thyroid cell irradiation by radioiodines: a new Monte Carlo electron track-structure code

    Directory of Open Access Journals (Sweden)

    Christophe Champion

    2007-09-01

    Full Text Available The most significant impact of the Chernobyl accident is the increased incidence of thyroid cancer among children who were exposed to short-lived radioiodines and 131-iodine. In order to accurately estimate the radiation dose provided by these radioiodines, it is necessary to know where iodine is incorporated. To do that, the distribution at the cellular level of newly organified iodine in the immature rat thyroid was performed using secondary ion mass microscopy (NanoSIMS50. Actual dosimetric models take only into account the averaged energy and range of beta particles of the radio-elements and may, therefore, imperfectly describe the real distribution of dose deposit at the microscopic level around the point sources. Our approach is radically different since based on a track-structure Monte Carlo code allowing following-up of electrons down to low energies (~ 10eV what permits a nanometric description of the irradiation physics. The numerical simulations were then performed by modelling the complete disintegrations of the short-lived iodine isotopes as well as of 131I in new born rat thyroids in order to take into account accurate histological and biological data for the thyroid gland.O impacto mais significante do acidente de Chernobyl é o crescimento da incidência de câncer de tireóide em crianças que foram expostas a radioiodos de vida curta e ao Iodo-131. Na estimativa precisa da dose de radiação fornecida por esses radioiodos, é necessário conhecer onde o iodo está incorporado. Para obtermos esse resultado, a distribuição em nível celular de iodo recentemente organificado na tireóde de ratos imaturos foi realizada usando microscopia de massa iônica secundária (NanoSIMS50. Modelos dosimétricos atuais consideram apenas a energia média das partículas beta dos radioelementos e pode, imperfeitamente descrever a distribuição real de dose ao nível microscópico em torno dos pontos pesquisados. Nossa abordagem

  15. A Monte Carlo Simulation for the Ion Transport in Glow Discharges with Dusts

    Institute of Scientific and Technical Information of China (English)

    SUN Ai-Ping; PU Wei; QIU Xiao-Ming

    2001-01-01

    We use the Monte Carlo method to simulate theion transport in the rf parallel plate glow discharge with a negative-voltage pulse connected to the electrode. It is found that self-consistent field, dust charge, dust concentration,and dust size influence the energy distribution and the density of the ions arriving at the target, and in particular, the latter two make significant influence. As dust concentration or dust size increases, the number of ions arriving at the target reduces greatly.

  16. FitSKIRT: genetic algorithms to automatically fit dusty galaxies with a Monte Carlo radiative transfer code

    CERN Document Server

    De Geyter, Gert; Fritz, Jacopo; Camps, Peter

    2012-01-01

    We present FitSKIRT, a method to efficiently fit radiative transfer models to UV/optical images of dusty galaxies. These images have the advantage that they have better spatial resolution compared to FIR/submm data. FitSKIRT uses the GAlib genetic algorithm library to optimize the output of the SKIRT Monte Carlo radiative transfer code. Genetic algorithms prove to be a valuable tool in handling the multi- dimensional search space as well as the noise induced by the random nature of the Monte Carlo radiative transfer code. FitSKIRT is tested on artificial images of a simulated edge-on spiral galaxy, where we gradually increase the number of fitted parameters. We find that we can recover all model parameters, even if all 11 model parameters are left unconstrained. Finally, we apply the FitSKIRT code to a V-band image of the edge-on spiral galaxy NGC4013. This galaxy has been modeled previously by other authors using different combinations of radiative transfer codes and optimization methods. Given the different...

  17. Improvements to the National Transport Code Collaboration Data Server

    Science.gov (United States)

    Alexander, David A.

    2001-10-01

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

  18. TRIGA IPR-R1 reactor simulation using Monte Carlo transport methods

    OpenAIRE

    Hugo Moura Dalle

    2005-01-01

    Resumo: A utilização do método Monte Carlo na simulação do transporte de partículas em reatores nucleares é crescente e constitui uma tendência mundial. O maior inconveniente dessa técnica, a grande exigência de capacidade de processamento, vem sendo superado pelo contínuo desenvolvimento de processadores cada vez mais rápidos. Esse contexto permitiu o desenvolvimento de metodologias de cálculo neutrônico de reatores nas quais se acopla a parte do transporte de partículas, feita com um código...

  19. A Deterministic-Monte Carlo Hybrid Method for Time-Dependent Neutron Transport Problems

    Energy Technology Data Exchange (ETDEWEB)

    Justin Pounders; Farzad Rahnema

    2001-10-01

    A new deterministic-Monte Carlo hybrid solution technique is derived for the time-dependent transport equation. This new approach is based on dividing the time domain into a number of coarse intervals and expanding the transport solution in a series of polynomials within each interval. The solutions within each interval can be represented in terms of arbitrary source terms by using precomputed response functions. In the current work, the time-dependent response function computations are performed using the Monte Carlo method, while the global time-step march is performed deterministically. This work extends previous work by coupling the time-dependent expansions to space- and angle-dependent expansions to fully characterize the 1D transport response/solution. More generally, this approach represents and incremental extension of the steady-state coarse-mesh transport method that is based on global-local decompositions of large neutron transport problems. An example of a homogeneous slab is discussed as an example of the new developments.

  20. Application of a Monte Carlo Penelope code at diverse dosimetric problems in radiotherapy; Aplicacion del codigo Monte Carlo Penelope a diversos problemas dosimetricos en radioterapia

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, R.A.; Fernandez V, J.M.; Salvat, F. [Servicio de Oncologia Radioterapica. Hospital Clinico de Barcelona. Villarroel 170 08036 Barcelona (Spain)

    1998-12-31

    In the present communication it is presented the results of the simulation utilizing the Penelope code (Penetration and Energy loss of Positrons and Electrons) in several applications of radiotherapy which can be the radioactive sources simulation: {sup 192} Ir, {sup 125} I, {sup 106} Ru or the electron beams simulation of a linear accelerator Siemens KDS. The simulations presented in this communication have been on computers of type Pentium PC of 100 throughout 300 MHz, and the times of execution were from some hours until several days depending of the complexity of the problem. It is concluded that Penelope is a very useful tool for the Monte Carlo calculations due to its great ability and its relative handling facilities. (Author)

  1. A new Monte Carlo code for simulation of the effect of irregular surfaces on X-ray spectra

    Energy Technology Data Exchange (ETDEWEB)

    Brunetti, Antonio, E-mail: brunetti@uniss.it; Golosio, Bruno

    2014-04-01

    Generally, quantitative X-ray fluorescence (XRF) analysis estimates the content of chemical elements in a sample based on the areas of the fluorescence peaks in the energy spectrum. Besides the concentration of the elements, the peak areas depend also on the geometrical conditions. In fact, the estimate of the peak areas is simple if the sample surface is smooth and if the spectrum shows a good statistic (large-area peaks). For this reason often the sample is prepared as a pellet. However, this approach is not always feasible, for instance when cultural heritage or valuable samples must be analyzed. In this case, the sample surface cannot be smoothed. In order to address this problem, several works have been reported in the literature, based on experimental measurements on a few sets of specific samples or on Monte Carlo simulations. The results obtained with the first approach are limited by the specific class of samples analyzed, while the second approach cannot be applied to arbitrarily irregular surfaces. The present work describes a more general analysis tool based on a new fast Monte Carlo algorithm, which is virtually able to simulate any kind of surface. At the best of our knowledge, it is the first Monte Carlo code with this option. A study of the influence of surface irregularities on the measured spectrum is performed and some results reported. - Highlights: • We present a fast Monte Carlo code with the possibility to simulate any irregularly rough surfaces. • We show applications to multilayer measurements. • Real time simulations are available.

  2. Benchmarking of the dose planning method (DPM) Monte Carlo code using electron beams from a racetrack microtron.

    Science.gov (United States)

    Chetty, Indrin J; Moran, Jean M; McShan, Daniel L; Fraass, Benedick A; Wilderman, Scott J; Bielajew, Alex F

    2002-06-01

    A comprehensive set of measurements and calculations has been conducted to investigate the accuracy of the Dose Planning Method (DPM) Monte Carlo code for dose calculations from 10 and 50 MeV scanned electron beams produced from a racetrack microtron. Central axis depth dose measurements and a series of profile scans at various depths were acquired in a water phantom using a Scanditronix type RK ion chamber. Source spatial distributions for the Monte Carlo calculations were reconstructed from in-air ion chamber measurements carried out across the two-dimensional beam profile at 100 cm downstream from the source. The in-air spatial distributions were found to have full width at half maximum of 4.7 and 1.3 cm, at 100 cm from the source, for the 10 and 50 MeV beams, respectively. Energy spectra for the 10 and 50 MeV beams were determined by simulating the components of the microtron treatment head using the code MCNP4B. DPM calculations are on average within +/- 2% agreement with measurement for all depth dose and profile comparisons conducted in this study. The accuracy of the DPM code illustrated in this work suggests that DPM may be used as a valuable tool for electron beam dose calculations.

  3. Non-thermodynamic approach to including bombardment-induced post-cascade redistribution of point defects in dynamic Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Ignatova, V.A. E-mail: velislav@uia.ua.ac.be; Chakarov, I.R.; Katardjiev, I.V

    2003-04-01

    The redistribution of the elements as a result of atomic relocations produced by the ions and the recoils due to the ballistic and transport processes is investigated by making use of a dynamic Monte Carlo code. Phenomena, such as radiation-enhanced diffusion (RED) and bombardment-induced segregation (BIS) triggered by the ion bombardment may also contribute to the migration of atoms within the target. In order to include both RED and BIS in the code, we suggest an approach which is considered as an extension of the binary collision approximation, i.e. it takes place 'simultaneously' with the cascade and acts as a correction to the particle redistribution for low energies. Both RED and BIS models are based on the common approach to treat the transport processes as a result of a random migration of point defects (vacancies and interstitials) according to a probability given by a pre-defined Gaussian. The models are tested and the influence of the diffusion and segregation is illustrated in the cases of 12 keV {sup 121}Sb{sup +} implantation at low fluence in SiO{sub 2}/Si substrate and of self-sputtering of Ga{sup +} ions during profiling of SiO{sub 2}/Si interfaces.

  4. Non-thermodynamic approach to including bombardment-induced post-cascade redistribution of point defects in dynamic Monte Carlo code

    CERN Document Server

    Ignatova, V A; Katardjiev, I V

    2003-01-01

    The redistribution of the elements as a result of atomic relocations produced by the ions and the recoils due to the ballistic and transport processes is investigated by making use of a dynamic Monte Carlo code. Phenomena, such as radiation-enhanced diffusion (RED) and bombardment-induced segregation (BIS) triggered by the ion bombardment may also contribute to the migration of atoms within the target. In order to include both RED and BIS in the code, we suggest an approach which is considered as an extension of the binary collision approximation, i.e. it takes place 'simultaneously' with the cascade and acts as a correction to the particle redistribution for low energies. Both RED and BIS models are based on the common approach to treat the transport processes as a result of a random migration of point defects (vacancies and interstitials) according to a probability given by a pre-defined Gaussian. The models are tested and the influence of the diffusion and segregation is illustrated in the cases of 12 keV ...

  5. 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.

  6. Development of perturbation Monte Carlo methods for polarized light transport in a discrete particle scattering model.

    Science.gov (United States)

    Nguyen, Jennifer; Hayakawa, Carole K; Mourant, Judith R; Venugopalan, Vasan; Spanier, Jerome

    2016-05-01

    We present a polarization-sensitive, transport-rigorous perturbation Monte Carlo (pMC) method to model the impact of optical property changes on reflectance measurements within a discrete particle scattering model. The model consists of three log-normally distributed populations of Mie scatterers that approximate biologically relevant cervical tissue properties. Our method provides reflectance estimates for perturbations across wavelength and/or scattering model parameters. We test our pMC model performance by perturbing across number densities and mean particle radii, and compare pMC reflectance estimates with those obtained from conventional Monte Carlo simulations. These tests allow us to explore different factors that control pMC performance and to evaluate the gains in computational efficiency that our pMC method provides.

  7. Numerical Study of Light Transport in Apple Models Based on Monte Carlo Simulations

    Directory of Open Access Journals (Sweden)

    Mohamed Lamine Askoura

    2015-12-01

    Full Text Available This paper reports on the quantification of light transport in apple models using Monte Carlo simulations. To this end, apple was modeled as a two-layer spherical model including skin and flesh bulk tissues. The optical properties of both tissue types used to generate Monte Carlo data were collected from the literature, and selected to cover a range of values related to three apple varieties. Two different imaging-tissue setups were simulated in order to show the role of the skin on steady-state backscattering images, spatially-resolved reflectance profiles, and assessment of flesh optical properties using an inverse nonlinear least squares fitting algorithm. Simulation results suggest that apple skin cannot be ignored when a Visible/Near-Infrared (Vis/NIR steady-state imaging setup is used for investigating quality attributes of apples. They also help to improve optical inspection techniques in the horticultural products.

  8. Differences among Monte Carlo codes in the calculations of voxel S values for radionuclide targeted therapy and analysis of their impact on absorbed dose evaluations

    Energy Technology Data Exchange (ETDEWEB)

    Pacilio, M.; Lanconelli, N.; Lo Meo, S.; Betti, M.; Montani, L.; Torres Aroche, L. A.; Coca Perez, M. A. [Department of Medical Physics, Azienda Ospedaliera S. Camillo Forlanini, Piazza Forlanini 1, Rome 00151 (Italy); Department of Physics, Alma Mater Studiorum University of Bologna, Viale Berti-Pichat 6/2, Bologna 40127 (Italy); Department of Medical Physics, Azienda Ospedaliera S. Camillo Forlanini, Piazza Forlanini 1, Rome 00151 (Italy); Department of Medical Physics, Azienda Ospedaliera Sant' Andrea, Via di Grotarossa 1035, Rome 00189 (Italy); Department of Medical Physics, Center for Clinical Researches, Calle 34 North 4501, Havana 11300 (Cuba)

    2009-05-15

    Several updated Monte Carlo (MC) codes are available to perform calculations of voxel S values for radionuclide targeted therapy. The aim of this work is to analyze the differences in the calculations obtained by different MC codes and their impact on absorbed dose evaluations performed by voxel dosimetry. Voxel S values for monoenergetic sources (electrons and photons) and different radionuclides ({sup 90}Y, {sup 131}I, and {sup 188}Re) were calculated. Simulations were performed in soft tissue. Three general-purpose MC codes were employed for simulating radiation transport: MCNP4C, EGSnrc, and GEANT4. The data published by the MIRD Committee in Pamphlet No. 17, obtained with the EGS4 MC code, were also included in the comparisons. The impact of the differences (in terms of voxel S values) among the MC codes was also studied by convolution calculations of the absorbed dose in a volume of interest. For uniform activity distribution of a given radionuclide, dose calculations were performed on spherical and elliptical volumes, varying the mass from 1 to 500 g. For simulations with monochromatic sources, differences for self-irradiation voxel S values were mostly confined within 10% for both photons and electrons, but with electron energy less than 500 keV, the voxel S values referred to the first neighbor voxels showed large differences (up to 130%, with respect to EGSnrc) among the updated MC codes. For radionuclide simulations, noticeable differences arose in voxel S values, especially in the bremsstrahlung tails, or when a high contribution from electrons with energy of less than 500 keV is involved. In particular, for {sup 90}Y the updated codes showed a remarkable divergence in the bremsstrahlung region (up to about 90% in terms of voxel S values) with respect to the EGS4 code. Further, variations were observed up to about 30%, for small source-target voxel distances, when low-energy electrons cover an important part of the emission spectrum of the radionuclide

  9. Experimental verification of NOVICE transport code predictions of electron distributions from targets

    CERN Document Server

    Kronenberg, S; Jordan, T; Bechtel, E; Gentner, F; Groeber, E

    2002-01-01

    This paper reports the results of experiments that were designed to check the validity of the NOVICE Adjoint Monte Carlo Transport code in predicting emission-electron distributions from irradiated targets. Previous work demonstrated that the code accurately calculated total electron yields from irradiated targets. In this investigation, a gold target was irradiated by X-rays with effective quantum energies of 79, 127, 174, 216, and 250 keV. Spectra of electrons from the target were measured for an incident photon angle of 45 deg., an emission-electron polar angle of 45 deg., azimuthal angles of 0 deg. and 180 deg., and in both the forward and backward directions. NOVICE was used to predict those electron-energy-distributions for the same set of experimental conditions. The agreement in shape of the theoretical and experimental distributions was good, whereas the absolute agreement in amplitude was within about a factor of 2 over most of the energy range of the spectra. Previous experimental and theoretical c...

  10. Monte Carlo Study of Temperature-dependent Non-diffusive Thermal Transport in Si Nanowires

    CERN Document Server

    Ma, Lei; Liu, Mengmeng; Zhao, Xuxin; Wu, Qixing; Sun, Hongyuan

    2016-01-01

    Non-diffusive thermal transport has gained extensive research interest recently due to its important implications on fundamental understanding of material phonon mean free path distributions and many nanoscale energy applications. In this work, we systematically investigate the role of boundary scattering and nanowire length on the nondiffusive thermal transport in thin silicon nanowires by rigorously solving the phonon Boltzmann transport equation using a variance reduced Monte Carlo technique across a range of temperatures. The simulations use the complete phonon dispersion and spectral lifetime data obtained from first-principle density function theory calculations as input without any adjustable parameters. Our BTE simulation results show that the nanowire length plays an important role in determining the thermal conductivity of silicon nanowires. In addition, our simulation results suggest significant phonon confinement effect for the previously measured silicon nanowires. These findings are important fo...

  11. Gamma ray transport simulations using SGaRD code

    Directory of Open Access Journals (Sweden)

    Humbert Philippe

    2017-01-01

    Full Text Available SGaRD (Spectroscopy, Gamma rays, Rapid, Deterministic code is used for the fast calculation of the gamma-ray spectrum, produced by a spherical shielded source and measured by a detector. The photon source lines originate from the radioactive decay of the unstable isotopes. The leakage spectrum is separated in two parts: the uncollided component is transported by ray tracing, and the scattered component is calculated using a multigroup discrete ordinates method. The pulse height spectrum is then simulated by folding the leakage spectrum with the detector response function, which is precalculated for each considered detector type. An application to the simulation of the gamma spectrum produced by a natural uranium ball coated with plexiglass and measured using a NaI detector is presented. The SGaRD code is also used to infer the dimensions of a one-dimensional model of a shielded gamma ray source. The method is based on the simulation of the uncollided leakage current of discrete gamma lines that are produced by nuclear decay. The material thicknesses are computed with SGaRD using a fast ray-tracing algorithm embedded in a nonlinear multidimensional iterative optimization procedure that minimizes the error metric between calculated and measured signatures.

  12. Development of NRESP98 Monte Carlo codes for the calculation of neutron response functions of neutron detectors. Calculation of the response function of spherical BF{sub 3} proportional counter

    Energy Technology Data Exchange (ETDEWEB)

    Hashimoto, M.; Saito, K.; Ando, H. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-05-01

    The method to calculate the response function of spherical BF{sub 3} proportional counter, which is commonly used as neutron dose rate meter and neutron spectrometer with multi moderator system, is developed. As the calculation code for evaluating the response function, the existing code series NRESP, the Monte Carlo code for the calculation of response function of neutron detectors, is selected. However, the application scope of the existing NRESP is restricted, the NRESP98 is tuned as generally applicable code, with expansion of the geometrical condition, the applicable element, etc. The NRESP98 is tested with the response function of the spherical BF{sub 3} proportional counter. Including the effect of the distribution of amplification factor, the detailed evaluation of the charged particle transportation and the effect of the statistical distribution, the result of NRESP98 calculation fit the experience within {+-}10%. (author)

  13. Final Report for National Transport Code Collaboration PTRANSP

    Energy Technology Data Exchange (ETDEWEB)

    Arnold H. Kritz

    2012-06-14

    PTRANSP, which is the predictive version of the TRANSP code, was developed in a collaborative effort involving the Princeton Plasma Physics Laboratory, General Atomics Corporation, Lawrence Livermore National Laboratory, and Lehigh University. The PTRANSP/TRANSP suite of codes is the premier integrated tokamak modeling software in the United States. A production service for PTRANSP/TRANSP simulations is maintained at the Princeton Plasma Physics Laboratory; the server has a simple command line client interface and is subscribed to by about 100 researchers from tokamak projects in the US, Europe, and Asia. This service produced nearly 13000 PTRANSP/TRANSP simulations in the four year period FY 2005 through FY 2008. Major archives of TRANSP results are maintained at PPPL, MIT, General Atomics, and JET. Recent utilization, counting experimental analysis simulations as well as predictive simulations, more than doubled from slightly over 2000 simulations per year in FY 2005 and FY 2006 to over 4300 simulations per year in FY 2007 and FY 2008. PTRANSP predictive simulations applied to ITER increased eight fold from 30 simulations per year in FY 2005 and FY 2006 to 240 simulations per year in FY 2007 and FY 2008, accounting for more than half of combined PTRANSP/TRANSP service CPU resource utilization in FY 2008. PTRANSP studies focused on ITER played a key role in journal articles. Examples of validation studies carried out for momentum transport in PTRANSP simulations were presented at the 2008 IAEA conference. The increase in number of PTRANSP simulations has continued (more than 7000 TRANSP/PTRANSP simulations in 2010) and results of PTRANSP simulations appear in conference proceedings, for example the 2010 IAEA conference, and in peer reviewed papers. PTRANSP provides a bridge to the Fusion Simulation Program (FSP) and to the future of integrated modeling. Through years of widespread usage, each of the many parts of the PTRANSP suite of codes has been thoroughly

  14. Hybrid Parallel Programming Models for AMR Neutron Monte-Carlo Transport

    Science.gov (United States)

    Dureau, David; Poëtte, Gaël

    2014-06-01

    This paper deals with High Performance Computing (HPC) applied to neutron transport theory on complex geometries, thanks to both an Adaptive Mesh Refinement (AMR) algorithm and a Monte-Carlo (MC) solver. Several Parallelism models are presented and analyzed in this context, among them shared memory and distributed memory ones such as Domain Replication and Domain Decomposition, together with Hybrid strategies. The study is illustrated by weak and strong scalability tests on complex benchmarks on several thousands of cores thanks to the petaflopic supercomputer Tera100.

  15. Microdosimetry of alpha particles for simple and 3D voxelised geometries using MCNPX and Geant4 Monte Carlo codes.

    Science.gov (United States)

    Elbast, M; Saudo, A; Franck, D; Petitot, F; Desbrée, A

    2012-07-01

    Microdosimetry using Monte Carlo simulation is a suitable technique to describe the stochastic nature of energy deposition by alpha particle at cellular level. Because of its short range, the energy imparted by this particle to the targets is highly non-uniform. Thus, to achieve accurate dosimetric results, the modelling of the geometry should be as realistic as possible. The objectives of the present study were to validate the use of the MCNPX and Geant4 Monte Carlo codes for microdosimetric studies using simple and three-dimensional voxelised geometry and to study their limit of validity in this last case. To that aim, the specific energy (z) deposited in the cell nucleus, the single-hit density of specific energy f(1)(z) and the mean-specific energy were calculated. Results show a good agreement when compared with the literature using simple geometry. The maximum percentage difference found is MCNPX for calculation time is 10 times higher with Geant4 than MCNPX code in the same conditions.

  16. MC3: Multi-core Markov-chain Monte Carlo code

    Science.gov (United States)

    Cubillos, Patricio; Harrington, Joseph; Lust, Nate; Foster, AJ; Stemm, Madison; Loredo, Tom; Stevenson, Kevin; Campo, Chris; Hardin, Matt; Hardy, Ryan

    2016-10-01

    MC3 (Multi-core Markov-chain Monte Carlo) is a Bayesian statistics tool that can be executed from the shell prompt or interactively through the Python interpreter with single- or multiple-CPU parallel computing. It offers Markov-chain Monte Carlo (MCMC) posterior-distribution sampling for several algorithms, Levenberg-Marquardt least-squares optimization, and uniform non-informative, Jeffreys non-informative, or Gaussian-informative priors. MC3 can share the same value among multiple parameters and fix the value of parameters to constant values, and offers Gelman-Rubin convergence testing and correlated-noise estimation with time-averaging or wavelet-based likelihood estimation methods.

  17. Monte Carlo Simulation of Siemens ONCOR Linear Accelerator with BEAMnrc and DOSXYZnrc Code.

    Science.gov (United States)

    Jabbari, Keyvan; Anvar, Hossein Saberi; Tavakoli, Mohammad Bagher; Amouheidari, Alireza

    2013-07-01

    The Monte Carlo method is the most accurate method for simulation of radiation therapy equipment. The linear accelerators (linac) are currently the most widely used machines in radiation therapy centers. In this work, a Monte Carlo modeling of the Siemens ONCOR linear accelerator in 6 MV and 18 MV beams was performed. The results of simulation were validated by measurements in water by ionization chamber and extended dose range (EDR2) film in solid water. The linac's X-ray particular are so sensitive to the properties of primary electron beam. Square field size of 10 cm × 10 cm produced by the jaws was compared with ionization chamber and film measurements. Head simulation was performed with BEAMnrc and dose calculation with DOSXYZnrc for film measurements and 3ddose file produced by DOSXYZnrc analyzed used homemade MATLAB program. At 6 MV, the agreement between dose calculated by Monte Carlo modeling and direct measurement was obtained to the least restrictive of 1%, even in the build-up region. At 18 MV, the agreement was obtained 1%, except for in the build-up region. In the build-up region, the difference was 1% at 6 MV and 2% at 18 MV. The mean difference between measurements and Monte Carlo simulation is very small in both of ONCOR X-ray energy. The results are highly accurate and can be used for many applications such as patient dose calculation in treatment planning and in studies that model this linac with small field size like intensity-modulated radiation therapy technique.

  18. Evaluation of PENFAST - A fast Monte Carlo code for dose calculations in photon and electron radiotherapy treatment planning

    Energy Technology Data Exchange (ETDEWEB)

    Habib, B.; Poumarede, B.; Tola, F.; Barthe, J. [CEA, LIST, Dept Technol Capteur et Signal, F-91191 Gif Sur Yvette, (France)

    2010-07-01

    The aim of the present study is to demonstrate the potential of accelerated dose calculations, using the fast Monte Carlo (MC) code referred to as PENFAST, rather than the conventional MC code PENELOPE, without losing accuracy in the computed dose. For this purpose, experimental measurements of dose distributions in homogeneous and inhomogeneous phantoms were compared with simulated results using both PENELOPE and PENFAST. The simulations and experiments were performed using a Saturne 43 linac operated at 12 MV (photons), and at 18 MeV (electrons). Pre-calculated phase space files (PSFs) were used as input data to both the PENELOPE and PENFAST dose simulations. Since depth-dose and dose profile comparisons between simulations and measurements in water were found to be in good agreement (within {+-} 1% to 1 mm), the PSF calculation is considered to have been validated. In addition, measured dose distributions were compared to simulated results in a set of clinically relevant, inhomogeneous phantoms, consisting of lung and bone heterogeneities in a water tank. In general, the PENFAST results agree to within a 1% to 1 mm difference with those produced by PENELOPE, and to within a 2% to 2 mm difference with measured values. Our study thus provides a pre-clinical validation of the PENFAST code. It also demonstrates that PENFAST provides accurate results for both photon and electron beams, equivalent to those obtained with PENELOPE. CPU time comparisons between both MC codes show that PENFAST is generally about 9-21 times faster than PENELOPE. (authors)

  19. The Monte Carlo code CSSE for the simulation of realistic thermal neutron sensor devices for Humanitarian Demining

    Energy Technology Data Exchange (ETDEWEB)

    Palomba, M. E-mail: maurizio.palomba@ba.infn.it; D' Erasmo, G.; Pantaleo, A

    2003-02-11

    The CSSE code, a GEANT3-based Monte Carlo simulation program, has been developed in the framework of the EXPLODET project (Nucl. Instr. and Meth. A 422 (1999) 918) with the aim to simulate experimental set-ups employed in Thermal Neutron Analysis (TNA) for the landmines detection. Such a simulation code appears to be useful for studying the background in the {gamma}-ray spectra obtained with this technique, especially in the region where one expects to find the explosive signature (the {gamma}-ray peak at 10.83 MeV coming from neutron capture by nitrogen). The main features of the CSSE code are introduced and original innovations emphasized. Among the latter, an algorithm simulating the time correlation between primary particles, according with their time distributions is presented. Such a correlation is not usually achievable within standard GEANT-based codes and allows to reproduce some important phenomena, as the pulse pile-up inside the NaI(Tl) {gamma}-ray detector employed, producing a more realistic detector response simulation. CSSE has been successfully tested by reproducing a real nuclear sensor prototype assembled at the Physics Department of Bari University.

  20. The Monte Carlo code CSSE for the simulation of realistic thermal neutron sensor devices for Humanitarian Demining

    Science.gov (United States)

    Palomba, M.; D'Erasmo, G.; Pantaleo, A.

    2003-02-01

    The CSSE code, a GEANT3-based Monte Carlo simulation program, has been developed in the framework of the EXPLODET project (Nucl. Instr. and Meth. A 422 (1999) 918) with the aim to simulate experimental set-ups employed in Thermal Neutron Analysis (TNA) for the landmines detection. Such a simulation code appears to be useful for studying the background in the γ-ray spectra obtained with this technique, especially in the region where one expects to find the explosive signature (the γ-ray peak at 10.83 MeV coming from neutron capture by nitrogen). The main features of the CSSE code are introduced and original innovations emphasized. Among the latter, an algorithm simulating the time correlation between primary particles, according with their time distributions is presented. Such a correlation is not usually achievable within standard GEANT-based codes and allows to reproduce some important phenomena, as the pulse pile-up inside the NaI(Tl) γ-ray detector employed, producing a more realistic detector response simulation. CSSE has been successfully tested by reproducing a real nuclear sensor prototype assembled at the Physics Department of Bari University.

  1. GTNEUT: A code for the calculation of neutral particle transport in plasmas based on the Transmission and Escape Probability method

    Science.gov (United States)

    Mandrekas, John

    2004-08-01

    GTNEUT is a two-dimensional code for the calculation of the transport of neutral particles in fusion plasmas. It is based on the Transmission and Escape Probabilities (TEP) method and can be considered a computationally efficient alternative to traditional Monte Carlo methods. The code has been benchmarked extensively against Monte Carlo and has been used to model the distribution of neutrals in fusion experiments. Program summaryTitle of program: GTNEUT Catalogue identifier: ADTX Program obtainable from: CPC Program Library, Queen's University of Belfast, N. Ireland Program summary URL:http://cpc.cs.qub.ac.uk/summaries/ADTX Computer for which the program is designed and others on which it has been tested: The program was developed on a SUN Ultra 10 workstation and has been tested on other Unix workstations and PCs. Operating systems or monitors under which the program has been tested: Solaris 8, 9, HP-UX 11i, Linux Red Hat v8.0, Windows NT/2000/XP. Programming language used: Fortran 77 Memory required to execute with typical data: 6 219 388 bytes No. of bits in a word: 32 No. of processors used: 1 Has the code been vectorized or parallelized?: No No. of bytes in distributed program, including test data, etc.: 300 709 No. of lines in distributed program, including test data, etc.: 17 365 Distribution format: compressed tar gzip file Keywords: Neutral transport in plasmas, Escape probability methods Nature of physical problem: This code calculates the transport of neutral particles in thermonuclear plasmas in two-dimensional geometric configurations. Method of solution: The code is based on the Transmission and Escape Probability (TEP) methodology [1], which is part of the family of integral transport methods for neutral particles and neutrons. The resulting linear system of equations is solved by standard direct linear system solvers (sparse and non-sparse versions are included). Restrictions on the complexity of the problem: The current version of the code can

  2. Icarus: A 2D direct simulation Monte Carlo (DSMC) code for parallel computers. User`s manual - V.3.0

    Energy Technology Data Exchange (ETDEWEB)

    Bartel, T.; Plimpton, S.; Johannes, J.; Payne, J.

    1996-10-01

    Icarus is a 2D Direct Simulation Monte Carlo (DSMC) code which has been optimized for the parallel computing environment. The code is based on the DSMC method of Bird and models from free-molecular to continuum flowfields in either cartesian (x, y) or axisymmetric (z, r) coordinates. Computational particles, representing a given number of molecules or atoms, are tracked as they have collisions with other particles or surfaces. Multiple species, internal energy modes (rotation and vibration), chemistry, and ion transport are modelled. A new trace species methodology for collisions and chemistry is used to obtain statistics for small species concentrations. Gas phase chemistry is modelled using steric factors derived from Arrhenius reaction rates. Surface chemistry is modelled with surface reaction probabilities. The electron number density is either a fixed external generated field or determined using a local charge neutrality assumption. Ion chemistry is modelled with electron impact chemistry rates and charge exchange reactions. Coulomb collision cross-sections are used instead of Variable Hard Sphere values for ion-ion interactions. The electrostatic fields can either be externally input or internally generated using a Langmuir-Tonks model. The Icarus software package includes the grid generation, parallel processor decomposition, postprocessing, and restart software. The commercial graphics package, Tecplot, is used for graphics display. The majority of the software packages are written in standard Fortran.

  3. Particle Communication and Domain Neighbor Coupling: Scalable Domain Decomposed Algorithms for Monte Carlo Particle Transport

    Energy Technology Data Exchange (ETDEWEB)

    O' Brien, M. J.; Brantley, P. S.

    2015-01-20

    In order to run Monte Carlo particle transport calculations on new supercomputers with hundreds of thousands or millions of processors, care must be taken to implement scalable algorithms. This means that the algorithms must continue to perform well as the processor count increases. In this paper, we examine the scalability of:(1) globally resolving the particle locations on the correct processor, (2) deciding that particle streaming communication has finished, and (3) efficiently coupling neighbor domains together with different replication levels. We have run domain decomposed Monte Carlo particle transport on up to 221 = 2,097,152 MPI processes on the IBM BG/Q Sequoia supercomputer and observed scalable results that agree with our theoretical predictions. These calculations were carefully constructed to have the same amount of work on every processor, i.e. the calculation is already load balanced. We also examine load imbalanced calculations where each domain’s replication level is proportional to its particle workload. In this case we show how to efficiently couple together adjacent domains to maintain within workgroup load balance and minimize memory usage.

  4. ASDEX Upgrade Edge Transport Studies by Turbulence and Braginskii Divertor Transport Codes

    Energy Technology Data Exchange (ETDEWEB)

    Nishimura, Y.; Coster, D.P.; Kim, J.W.; Scott, B.D. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany)

    2001-07-01

    The equilibration time for diverter transport simulations is in the range of milliseconds to seconds. There, perpendicular transport is given empirically and usually assumed to be constant in time and space. In this work, we aim at describing edge plasma profiles in both the H-mode and the L-mode confinement regimes using a model that couples the transport scale to the underlying turbulence scale. There are 2d and 3d variants of DALF, which is a turbulence code that describes short time scale nonlinear phenomena based on first principles of plasma physics. B2 employs an implicit method which is suitable for describing long time scale, quasi-steady state behavior, while fluctuation/intermittency is inherent in turbulence and typically gives rise to short time scale variations of the radial flux. We coarse rained the information from the 2d version of DALF within the order of turbulence auto correlation time and iterated over the divertor simulation (and thus passed plasma parameters to the turbulence code). Numerical algorithm and criteria for convergence in bridging the physics of two different scales is discussed. The generation mechanism of radial electric field in steep gradient regimes is revisited in the ASDEX Upgrade divertor geometry with realistic parameters. Inclusion of turbulent suppression effects by E x B shear flow is considered. (orig.)

  5. Comparative Dosimetric Estimates of a 25 keV Electron Micro-beam with three Monte Carlo Codes

    CERN Document Server

    Mainardi, E; Donahue, R J

    2002-01-01

    The calculations presented compare the different performances of the three Monte Carlo codes PENELOPE-1999, MCNP-4C and PITS, for the evaluation of Dose profiles from a 25 keV electron micro-beam traversing individual cells. The overall model of a cell is a water cylinder equivalent for the three codes but with a different internal scoring geometry: hollow cylinders for PENELOPE and MCNP, whereas spheres are used for the PITS code. A cylindrical cell geometry with scoring volumes with the shape of hollow cylinders was initially selected for PENELOPE and MCNP because of its superior simulation of the actual shape and dimensions of a cell and for its improved computer-time efficiency if compared to spherical internal volumes. Some of the transfer points and energy transfer that constitute a radiation track may actually fall in the space between spheres, that would be outside the spherical scoring volume. This internal geometry, along with the PENELOPE algorithm, drastically reduced the computer time when using ...

  6. An integrated high-performance beam optics-nuclear processes framework with hybrid transfer map-Monte Carlo particle transport and optimization

    Energy Technology Data Exchange (ETDEWEB)

    Bandura, L., E-mail: bandura@msu.ed [Argonne National Laboratory, Argonne, IL 60439 (United States); Erdelyi, B. [Argonne National Laboratory, Argonne, IL 60439 (United States); Northern Illinois University, DeKalb, IL 60115 (United States); Nolen, J. [Argonne National Laboratory, Argonne, IL 60439 (United States)

    2010-12-01

    An integrated beam optics-nuclear processes framework is essential for accurate simulation of fragment separator beam dynamics. The code COSY INFINITY provides powerful differential algebraic methods for modeling and beam dynamics simulations in absence of beam-material interactions. However, these interactions are key for accurately simulating the dynamics of heavy ion fragmentation and fission. We have developed an extended version of the code that includes these interactions, and a set of new tools that allow efficient and accurate particle transport: by transfer map in vacuum and by Monte Carlo methods in materials. The new framework is presented, along with several examples from a preliminary layout of a fragment separator for a facility for rare isotope beams.

  7. Recommended direct simulation Monte Carlo collision model parameters for modeling ionized air transport processes

    Energy Technology Data Exchange (ETDEWEB)

    Swaminathan-Gopalan, Krishnan; Stephani, Kelly A., E-mail: ksteph@illinois.edu [Department of Mechanical Science and Engineering, University of Illinois at Urbana-Champaign, Urbana, Illinois 61801 (United States)

    2016-02-15

    A systematic approach for calibrating the direct simulation Monte Carlo (DSMC) collision model parameters to achieve consistency in the transport processes is presented. The DSMC collision cross section model parameters are calibrated for high temperature atmospheric conditions by matching the collision integrals from DSMC against ab initio based collision integrals that are currently employed in the Langley Aerothermodynamic Upwind Relaxation Algorithm (LAURA) and Data Parallel Line Relaxation (DPLR) high temperature computational fluid dynamics solvers. The DSMC parameter values are computed for the widely used Variable Hard Sphere (VHS) and the Variable Soft Sphere (VSS) models using the collision-specific pairing approach. The recommended best-fit VHS/VSS parameter values are provided over a temperature range of 1000-20 000 K for a thirteen-species ionized air mixture. Use of the VSS model is necessary to achieve consistency in transport processes of ionized gases. The agreement of the VSS model transport properties with the transport properties as determined by the ab initio collision integral fits was found to be within 6% in the entire temperature range, regardless of the composition of the mixture. The recommended model parameter values can be readily applied to any gas mixture involving binary collisional interactions between the chemical species presented for the specified temperature range.

  8. THE INVESTIGATION OF BURNUP CHARACTERISTICS USING THE SERPENT MONTE CARLO CODE FOR A SODIUM COOLED FAST REACTOR

    Directory of Open Access Journals (Sweden)

    MEHMET E. KORKMAZ

    2014-06-01

    Full Text Available In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR. The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1 and 233U fuel (fuel pin 2 and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method and TTA (Transmutation Trajectory Analysis method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff was between 0.964 and 0.954 and peaking factor is 1.88867.

  9. The investigation of burnup characteristics using the serpent Monte Carlo code for a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, Mehmet E.; Agar, Osman [Karamanoglu Mehmetbey University, Faculty of Kamil Oezdag Science, Karaman (Turkmenistan)

    2014-06-15

    In this research, we investigated the burnup characteristics and the conversion of fertile {sup 232}Th into fissile {sup 233}U in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning {sup 232}Th fuel (fuel pin 1) and {sup 233}U fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.

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

    Energy Technology Data Exchange (ETDEWEB)

    Gudowska, I.; Svensson, R. [Karolinska Inst. (Sweden). Dept. of Medical Radiation Physics]|[Huddinge Univ. Hospital, Stockholm (Sweden). Dept. of Medical Physics; Sorcini, B. [Karolinska Inst. (Sweden). Dept. of Medical Radiation Physics]|[Stockholm Univ. (Sweden)

    2001-07-01

    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.)

  11. Influence of chromatin condensation on the number of direct DSB damages induced by ions studied using a Monte Carlo code.

    Science.gov (United States)

    Dos Santos, M; Clairand, I; Gruel, G; Barquinero, J F; Incerti, S; Villagrasa, C

    2014-10-01

    The purpose of this work is to evaluate the influence of the chromatin condensation on the number of direct double-strand break (DSB) damages induced by ions. Two geometries of chromosome territories containing either condensed or decondensed chromatin were implemented as biological targets in the Geant4 Monte Carlo simulation code and proton and alpha irradiation was simulated using the Geant4-DNA processes. A DBSCAN algorithm was used in order to detect energy deposition clusters that could give rise to single-strand breaks or DSBs on the DNA molecule. The results of this study show an increase in the number and complexity of DNA DSBs in condensed chromatin when compared with decondensed chromatin.

  12. Initial validation of 4D-model for a clinical PET scanner using the Monte Carlo code gate

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Igor F.; Lima, Fernando R.A.; Gomes, Marcelo S., E-mail: falima@cnen.gov.b [Centro Regional de Ciencias Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil); Vieira, Jose W.; Pacheco, Ludimila M. [Instituto Federal de Educacao, Ciencia e Tecnologia (IFPE), Recife, PE (Brazil); Chaves, Rosa M. [Instituto de Radium e Supervoltagem Ivo Roesler, Recife, PE (Brazil)

    2011-07-01

    Building exposure computational models (ECM) of emission tomography (PET and SPECT) currently has several dedicated computing tools based on Monte Carlo techniques (SimSET, SORTEO, SIMIND, GATE). This paper is divided into two steps: (1) using the dedicated code GATE (Geant4 Application for Tomographic Emission) to build a 4D model (where the fourth dimension is the time) of a clinical PET scanner from General Electric, GE ADVANCE, simulating the geometric and electronic structures suitable for this scanner, as well as some phenomena 4D, for example, rotating gantry; (2) the next step is to evaluate the performance of the model built here in the reproduction of test noise equivalent count rate (NEC) based on the NEMA Standards Publication NU protocols 2-2007 for this tomography. The results for steps (1) and (2) will be compared with experimental and theoretical values of the literature showing actual state of art of validation. (author)

  13. Energy distribution of cosmic rays in the Earth’s atmosphere and avionic area using Monte Carlo codes

    Indian Academy of Sciences (India)

    MOHAMED M OULD; DIB A S A; BELBACHIR A H

    2016-07-01

    Cosmic rays cause significant damage to the electronic equipments of the aircrafts. In this paper, we have investigated the accumulation of the deposited energy of cosmic rays on the Earth’s atmosphere, especially in the aircraft area. In fact, if a high-energy neutron or proton interacts with a nanodevice having only a few atoms, this neutron or proton particle can change the nature of this device and destroy it. Our simulation based on Monte Carlo using Geant4 code shows that the deposited energy of neutron particles ranging between 200MeV and 5 GeV are strongly concentrated in the region between 10 and 15 km from the sea level which is exactly the avionic area. However, the Bragg peak energy of proton particle is slightly localized above the avionic area.

  14. Characterisation of the TRIUMF neutron facility using a Monte Carlo simulation code.

    Science.gov (United States)

    Monk, S D; Abram, T; Joyce, M J

    2015-04-01

    Here, the characterisation of the high-energy neutron field at TRIUMF (The Tri Universities Meson Facility, Vancouver, British Columbia) with Monte Carlo simulation software is described. The package used is MCNPX version 2.6.0, with the neutron fluence rate determined at three locations within the TRIUMF Thermal Neutron Facility (TNF), including the exit of the neutron channel where users of the facility can test devices that may be susceptible to the effects of this form of radiation. The facility is often used to roughly emulate the field likely to be encountered at high altitudes due to radiation of galactic origin and thus the simulated information is compared with the energy spectrum calculated to be due to neutron radiation of cosmic origin at typical aircraft altitudes. The calculated values were also compared with neutron flux measurements that were estimated using the activation of various foils by the staff of the facility, showing agreement within an order of magnitude.

  15. Comparative study among simulations of an internal monitoring system using different Monte Carlo codes; Estudo comparativo entre simulacoes de um sistema de monitoracao ocupacional interna utilizando diferentes codigos de Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Fonseca, T.C.F.; Bastos, F.M.; Figueiredo, M.T.T.; Souza, L.S.; Guimaraes, M.C.; Silva, C.R.E.; Mello, O.A.; Castelo e Silva, L.A.; Paixao, L., E-mail: tcff01@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Benavente, J.A.; Paiva, F.G. [Universidade Federal de Minas Gerais (PCTN/UFMG), Belo Horizonte, MG (Brazil). Curso de Pos-Graduacao em Ciencias e Tecnicas Nucleares

    2015-07-01

    Computational Monte Carlo (MC) codes have been used for simulation of nuclear installations mainly for internal monitoring of workers, the well known as Whole Body Counters (WBC). The main goal of this project was the modeling and simulation of the counting efficiency (CE) of a WBC system using three different MC codes: MCNPX, EGSnrc and VMC in-vivo. The simulations were performed for three different groups of analysts. The results shown differences between the three codes, as well as in the results obtained by the same code and modeled by different analysts. Moreover, all the results were also compared to the experimental results obtained in laboratory for meaning of validation and final comparison. In conclusion, it was possible to detect the influence on the results when the system is modeled by different analysts using the same MC code and in which MC code the results were best suited, when comparing to the experimental data result. (author)

  16. COMET-PE as an Alternative to Monte Carlo for Photon and Electron Transport

    Science.gov (United States)

    Hayward, Robert M.; Rahnema, Farzad

    2014-06-01

    Monte Carlo methods are a central component of radiotherapy treatment planning, shielding design, detector modeling, and other applications. Long calculation times, however, can limit the usefulness of these purely stochastic methods. The coarse mesh method for photon and electron transport (COMET-PE) provides an attractive alternative. By combining stochastic pre-computation with a deterministic solver, COMET-PE achieves accuracy comparable to Monte Carlo methods in only a fraction of the time. The method's implementation has been extended to 3D, and in this work, it is validated by comparison to DOSXYZnrc using a photon radiotherapy benchmark. The comparison demonstrates excellent agreement; of the voxels that received more than 10% of the maximum dose, over 97.3% pass a 2% / 2mm acceptance test and over 99.7% pass a 3% / 3mm test. Furthermore, the method is over an order of magnitude faster than DOSXYZnrc and is able to take advantage of both distributed-memory and shared-memory parallel architectures for increased performance.

  17. Comparison of experimental and Monte-Carlo simulation of MeV particle transport through tapered/straight glass capillaries and circular collimators

    Energy Technology Data Exchange (ETDEWEB)

    Hespeels, F., E-mail: felicien.hespeels@unamur.be [University of Namur, PMR, 61 rue de Bruxelles, 5000 Namur (Belgium); Tonneau, R. [University of Namur, PMR, 61 rue de Bruxelles, 5000 Namur (Belgium); Ikeda, T. [RIKEN Nishina Center, 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Lucas, S. [University of Namur, PMR, 61 rue de Bruxelles, 5000 Namur (Belgium)

    2015-11-01

    Highlights: • Monte-Carlo simulation for beam transportation through collimations devices. • We confirm the focusing effect of tapered glass capillary. • We confirm the feasibility of using passive collimation devices for ion beam analysis application. - Abstract: This study compares the capabilities of three different passive collimation devices to produce micrometer-sized beams for proton and alpha particle beams (1.7 MeV and 5.3 MeV respectively): classical platinum TEM-like collimators, straight glass capillaries and tapered glass capillaries. In addition, we developed a Monte-Carlo code, based on the Rutherford scattering theory, which simulates particle transportation through collimating devices. The simulation results match the experimental observations of beam transportation through collimators both in air and vacuum. This research shows the focusing effects of tapered capillaries which clearly enable higher transmission flux. Nevertheless, the capillaries alignment with an incident beam is a prerequisite but is tedious, which makes the TEM collimator the easiest way to produce a 50 μm microbeam.

  18. Monte Carlo Simulations of Charge Transport in 2D Organic Photovoltaics.

    Science.gov (United States)

    Gagorik, Adam G; Mohin, Jacob W; Kowalewski, Tomasz; Hutchison, Geoffrey R

    2013-01-01

    The effect of morphology on charge transport in organic photovoltaics is assessed using Monte Carlo. In isotopic two-phase morphologies, increasing the domain size from 6.3 to 18.3 nm improves the fill factor by 11.6%, a result of decreased tortuosity and relaxation of Coulombic barriers. Additionally, when small aggregates of electron acceptors are interdispersed into the electron donor phase, charged defects form in the system, reducing fill factors by 23.3% on average, compared with systems without aggregates. In contrast, systems with idealized connectivity show a 3.31% decrease in fill factor when domain size was increased from 4 to 64 nm. We attribute this to a decreased rate of exciton separation at donor-acceptor interfaces. Finally, we notice that the presence of Coulomb interactions increases device performance as devices become smaller. The results suggest that for commonly found isotropic morphologies the Coulomb interactions between charge carriers dominates exciton separation effects.

  19. Monte Carlo Simulations of Spin Transport in Nanoscale InGaAs Field Effect Transistors

    CERN Document Server

    Thorpe, B; Langbein, F; Schirmer, S

    2016-01-01

    By augmenting an in-house developed, experimentally verified Monte Carlo device simulator with a Bloch equation model with a spin-orbit interaction Hamiltonian accounting for Dresselhaus and Rashba couplings, we simulate electron spin transport in a \\SI{25}{nm} gate length InGaAs MOSFET. We observe non-uniform decay of the net magnetization between the source and gate electrodes and an interesting magnetization recovery effect due to spin refocusing induced by high electric field between the gate and drain electrodes. We demonstrate coherent control of the polarization vector of the drain current via the source-drain and gate voltages, and show that the magnetization of the drain current is sensitive to strain in the channel, suggesting that the device could act as a room-temperature nanoscale strain sensor.

  20. GPU-based high performance Monte Carlo simulation in neutron transport

    Energy Technology Data Exchange (ETDEWEB)

    Heimlich, Adino; Mol, Antonio C.A.; Pereira, Claudio M.N.A. [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Inteligencia Artificial Aplicada], e-mail: cmnap@ien.gov.br

    2009-07-01

    Graphics Processing Units (GPU) are high performance co-processors intended, originally, to improve the use and quality of computer graphics applications. Since researchers and practitioners realized the potential of using GPU for general purpose, their application has been extended to other fields out of computer graphics scope. The main objective of this work is to evaluate the impact of using GPU in neutron transport simulation by Monte Carlo method. To accomplish that, GPU- and CPU-based (single and multicore) approaches were developed and applied to a simple, but time-consuming problem. Comparisons demonstrated that the GPU-based approach is about 15 times faster than a parallel 8-core CPU-based approach also developed in this work. (author)

  1. Core-scale solute transport model selection using Monte Carlo analysis

    CERN Document Server

    Malama, Bwalya; James, Scott C

    2013-01-01

    Model applicability to core-scale solute transport is evaluated using breakthrough data from column experiments conducted with conservative tracers tritium (H-3) and sodium-22, and the retarding solute uranium-232. The three models considered are single-porosity, double-porosity with single-rate mobile-immobile mass-exchange, and the multirate model, which is a deterministic model that admits the statistics of a random mobile-immobile mass-exchange rate coefficient. The experiments were conducted on intact Culebra Dolomite core samples. Previously, data were analyzed using single- and double-porosity models although the Culebra Dolomite is known to possess multiple types and scales of porosity, and to exhibit multirate mobile-immobile-domain mass transfer characteristics at field scale. The data are reanalyzed here and null-space Monte Carlo analysis is used to facilitate objective model selection. Prediction (or residual) bias is adopted as a measure of the model structural error. The analysis clearly shows ...

  2. Monte Carlo simulation of phonon transport in variable cross-section nanowires

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    A dedicated Monte Carlo (MC) model is proposed to investigate the mechanism of phonon transport in variable cross-section silicon nanowires (NWs). Emphasis is placed on understanding the thermal rectification effect and thermal conduction in tapered cross-section and incremental cross-section NWs. In the simulations, both equal and unequal heat input conditions are discussed. Under the latter condition, the tapered cross-section NW has a more prominent thermal rectification effect. Additionally, the capacity of heat conduction in the tapered cross-section NW is always higher than that of the incremental one. Two reasons may be attributed to these behaviors: one is their different boundary conditions and the other is their different volume distribution. Although boundary scattering plays an important role in nanoscale structures, the results suggest the influence of boundary scattering on heat conduction is less obvious than that of volume distribution in NWs with variable cross-sections.

  3. Comparison of implicit and symbolic implicit Monte Carlo line transport with frequency weight vector extension

    Science.gov (United States)

    McKinley, Michael Scott; Brooks, Eugene D., III; Szoke, Abraham

    2003-07-01

    We compare the implicit Monte Carlo (IMC) technique to the symbolic IMC (SIMC) technique, with and without weight vectors in frequency space, for time-dependent line transport in the presence of collisional pumping. We examine the efficiency and accuracy of the IMC and SIMC methods for test problems involving the evolution of a collisionally pumped trapping problem to its steady-state, the surface heating of a cold medium by a beam, and the diffusion of energy from a localized region that is collisionally pumped. The importance of spatial biasing and teleportation for problems involving high opacity is demonstrated. Our numerical solution, along with its associated teleportation error, is checked against theoretical calculations for the last example.

  4. 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.

  5. Modeling parameterized geometry in GPU-based Monte Carlo particle transport simulation for radiotherapy.

    Science.gov (United States)

    Chi, Yujie; Tian, Zhen; Jia, Xun

    2016-08-01

    Monte Carlo (MC) particle transport simulation on a graphics-processing unit (GPU) platform has been extensively studied recently due to the efficiency advantage achieved via massive parallelization. Almost all of the existing GPU-based MC packages were developed for voxelized geometry. This limited application scope of these packages. The purpose of this paper is to develop a module to model parametric geometry and integrate it in GPU-based MC simulations. In our module, each continuous region was defined by its bounding surfaces that were parameterized by quadratic functions. Particle navigation functions in this geometry were developed. The module was incorporated to two previously developed GPU-based MC packages and was tested in two example problems: (1) low energy photon transport simulation in a brachytherapy case with a shielded cylinder applicator and (2) MeV coupled photon/electron transport simulation in a phantom containing several inserts of different shapes. In both cases, the calculated dose distributions agreed well with those calculated in the corresponding voxelized geometry. The averaged dose differences were 1.03% and 0.29%, respectively. We also used the developed package to perform simulations of a Varian VS 2000 brachytherapy source and generated a phase-space file. The computation time under the parameterized geometry depended on the memory location storing the geometry data. When the data was stored in GPU's shared memory, the highest computational speed was achieved. Incorporation of parameterized geometry yielded a computation time that was ~3 times of that in the corresponding voxelized geometry. We also developed a strategy to use an auxiliary index array to reduce frequency of geometry calculations and hence improve efficiency. With this strategy, the computational time ranged in 1.75-2.03 times of the voxelized geometry for coupled photon/electron transport depending on the voxel dimension of the auxiliary index array, and in 0

  6. Modeling parameterized geometry in GPU-based Monte Carlo particle transport simulation for radiotherapy

    Science.gov (United States)

    Chi, Yujie; Tian, Zhen; Jia, Xun

    2016-08-01

    Monte Carlo (MC) particle transport simulation on a graphics-processing unit (GPU) platform has been extensively studied recently due to the efficiency advantage achieved via massive parallelization. Almost all of the existing GPU-based MC packages were developed for voxelized geometry. This limited application scope of these packages. The purpose of this paper is to develop a module to model parametric geometry and integrate it in GPU-based MC simulations. In our module, each continuous region was defined by its bounding surfaces that were parameterized by quadratic functions. Particle navigation functions in this geometry were developed. The module was incorporated to two previously developed GPU-based MC packages and was tested in two example problems: (1) low energy photon transport simulation in a brachytherapy case with a shielded cylinder applicator and (2) MeV coupled photon/electron transport simulation in a phantom containing several inserts of different shapes. In both cases, the calculated dose distributions agreed well with those calculated in the corresponding voxelized geometry. The averaged dose differences were 1.03% and 0.29%, respectively. We also used the developed package to perform simulations of a Varian VS 2000 brachytherapy source and generated a phase-space file. The computation time under the parameterized geometry depended on the memory location storing the geometry data. When the data was stored in GPU’s shared memory, the highest computational speed was achieved. Incorporation of parameterized geometry yielded a computation time that was ~3 times of that in the corresponding voxelized geometry. We also developed a strategy to use an auxiliary index array to reduce frequency of geometry calculations and hence improve efficiency. With this strategy, the computational time ranged in 1.75-2.03 times of the voxelized geometry for coupled photon/electron transport depending on the voxel dimension of the auxiliary index array, and in 0

  7. Monte Carlo simulation of a multi-leaf collimator design for telecobalt machine using BEAMnrc code

    Directory of Open Access Journals (Sweden)

    Ayyangar Komanduri

    2010-01-01

    Full Text Available This investigation aims to design a practical multi-leaf collimator (MLC system for the cobalt teletherapy machine and check its radiation properties using the Monte Carlo (MC method. The cobalt machine was modeled using the BEAMnrc Omega-Beam MC system, which could be freely downloaded from the website of the National Research Council (NRC, Canada. Comparison with standard depth dose data tables and the theoretically modeled beam showed good agreement within 2%. An MLC design with low melting point alloy (LMPA was tested for leakage properties of leaves. The LMPA leaves with a width of 7 mm and height of 6 cm, with tongue and groove of size 2 mm wide by 4 cm height, produced only 4% extra leakage compared to 10 cm height tungsten leaves. With finite 60 Co source size, the interleaf leakage was insignificant. This analysis helped to design a prototype MLC as an accessory mount on a cobalt machine. The complete details of the simulation process and analysis of results are discussed.

  8. Burnup simulations of different fuel grades using the MCNPX Monte Carlo code

    Directory of Open Access Journals (Sweden)

    Asah-Opoku Fiifi

    2014-01-01

    Full Text Available Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX, uranium oxide fuel (UOX, and commercially enriched uranium or uranium metal (CEU - are used in this simulation and their impact on the effective multiplication factor (Keff and, hence, criticality and total radioactivity of the reactor core after fuel burnup analyzed. The effect of different clad materials on Keff is also studied. Burnup calculation results indicate a buildup of plutonium isotopes in UOX and CEU, as opposed to a decline in plutonium radioisotopes for MOX fuel burnup time. For MOX fuel, a decrease of 31.9% of the fissile plutonium isotope is observed, while for UOX and CEU, fissile plutonium isotopes increased by 82.3% and 83.8%, respectively. Keff results show zircaloy as a much more effective clad material in comparison to zirconium and stainless steel.

  9. Time dependent simulations of multiwavelength variability of the blazar Mrk 421 with a Monte Carlo multi-zone code

    CERN Document Server

    Chen, Xuhui; Liang, Edison; Boettcher, Markus

    2011-01-01

    (abridged) We present a new time-dependent multi-zone radiative transfer code and its application to study the SSC emission of Mrk 421. The code couples Fokker-Planck and Monte Carlo methods, in a 2D geometry. For the first time all the light travel time effects (LCTE) are fully considered, along with a proper treatment of Compton cooling, which depends on them. We study a set of simple scenarios where the variability is produced by injection of relativistic electrons as a `shock front' crosses the emission region. We consider emission from two components, with the second one either being pre-existing and co-spatial and participating in the evolution of the active region, or spatially separated and independent, only diluting the observed variability. Temporal and spectral results of the simulation are compared to the multiwavelength observations of Mrk 421 in March 2001. We find parameters that can adequately fit the observed SEDs and multiwavelength light curves and correlations. There remain however a few o...

  10. The FLUKA Monte Carlo code coupled with the local effect model for biological calculations in carbon ion therapy

    CERN Document Server

    Mairani, A; Kraemer, M; Sommerer, F; Parodi, K; Scholz, M; Cerutti, F; Ferrari, A; Fasso, A

    2010-01-01

    Clinical Monte Carlo (MC) calculations for carbon ion therapy have to provide absorbed and RBE-weighted dose. The latter is defined as the product of the dose and the relative biological effectiveness (RBE). At the GSI Helmholtzzentrum fur Schwerionenforschung as well as at the Heidelberg Ion Therapy Center (HIT), the RBE values are calculated according to the local effect model (LEM). In this paper, we describe the approach followed for coupling the FLUKA MC code with the LEM and its application to dose and RBE-weighted dose calculations for a superimposition of two opposed C-12 ion fields as applied in therapeutic irradiations. The obtained results are compared with the available experimental data of CHO (Chinese hamster ovary) cell survival and the outcomes of the GSI analytical treatment planning code TRiP98. Some discrepancies have been observed between the analytical and MC calculations of absorbed physical dose profiles, which can be explained by the differences between the laterally integrated depth-d...

  11. Feasibility study of photo-neutron flux in various irradiation channels of Ghana MNSR using a Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Birikorang, S.A., E-mail: anddydat@yahoo.com [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); Akaho, E.H.K.; Nyarko, B.J.B. [National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra-Ghana (Ghana); Ampomah-Amoako, E.; Seth, Debrah K.; Gyabour, R.A.; Sogbgaji, R.B.M. [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana)

    2011-07-15

    Highlights: > The photo-neutron source was investigated within Ghana MNSR irradiation channels. > Irradiation channels under study were inner, outer and the fission chamber. > Thermal rated power at sub-critical state was estimated. > Neutron flux variation was investigated within the channels. > MCNP code has been used to investigate the flux variation. - Abstract: Computer simulation was carried out for photo-neutron source variation in outer irradiation channel, inner irradiation channels and the fission channel of a tank-in-pool reactor, a Miniature Neutron Source Reactor (MNSR) in sub-critical condition. Evaluation of the photo-neutron was done after the reactor has been in sub-critical condition for three month period using Monte Carlo Neutron Particle (MCNP) code. Neutron flux monitoring from the Micro Computer Control Loop System (MCCLS) was also investigated at sub-critical condition. The recorded neutron fluxes from the MCCLS after investigations were used to calculate the power of the reactor at sub-critical state. The computed power at sub-critical state was used to normalize the un-normalized results from the MCNP.

  12. Assessment of ocular beta radiation dose distribution due to 106Ru/106Rh brachytherapy applicators using MCNPX Monte Carlo code

    Directory of Open Access Journals (Sweden)

    Nilseia Aparecida Barbosa

    2014-08-01

    Full Text Available Purpose: Melanoma at the choroid region is the most common primary cancer that affects the eye in adult patients. Concave ophthalmic applicators with 106Ru/106Rh beta sources are the more used for treatment of these eye lesions, mainly lesions with small and medium dimensions. The available treatment planning system for 106Ru applicators is based on dose distributions on a homogeneous water sphere eye model, resulting in a lack of data in the literature of dose distributions in the eye radiosensitive structures, information that may be crucial to improve the treatment planning process, aiming the maintenance of visual acuity. Methods: The Monte Carlo code MCNPX was used to calculate the dose distribution in a complete mathematical model of the human eye containing a choroid melanoma; considering the eye actual dimensions and its various component structures, due to an ophthalmic brachytherapy treatment, using 106Ru/106Rh beta-ray sources. Two possibilities were analyzed; a simple water eye and a heterogeneous eye considering all its structures. Two concave applicators, CCA and CCB manufactured by BEBIG and a complete mathematical model of the human eye were modeled using the MCNPX code. Results and Conclusion: For both eye models, namely water model and heterogeneous model, mean dose values simulated for the same eye regions are, in general, very similar, excepting for regions very distant from the applicator, where mean dose values are very low, uncertainties are higher and relative differences may reach 20.4%. For the tumor base and the eye structures closest to the applicator, such as sclera, choroid and retina, the maximum difference observed was 4%, presenting the heterogeneous model higher mean dose values. For the other eye regions, the higher doses were obtained when the homogeneous water eye model is taken into consideration. Mean dose distributions determined for the homogeneous water eye model are similar to those obtained for the

  13. Development of Momentum Conserving Monte Carlo Simulation Code for ECCD Study in Helical Plasmas

    Directory of Open Access Journals (Sweden)

    Murakami S.

    2015-01-01

    Full Text Available Parallel momentum conserving collision model is developed for GNET code, in which a linearized drift kinetic equation is solved in the five dimensional phase-space to study the electron cyclotron current drive (ECCD in helical plasmas. In order to conserve the parallel momentum, we introduce a field particle collision term in addition to the test particle collision term. Two types of the field particle collision term are considered. One is the high speed limit model, where the momentum conserving term does not depend on the velocity of the background plasma and can be expressed in a simple form. The other is the velocity dependent model, which is derived from the Fokker–Planck collision term directly. In the velocity dependent model the field particle operator can be expressed using Legendre polynominals and, introducing the Rosenbluth potential, we derive the field particle term for each Legendre polynominals. In the GNET code, we introduce an iterative process to implement the momentum conserving collision operator. The high speed limit model is applied to the ECCD simulation of the heliotron-J plasma. The simulation results show a good conservation of the momentum with the iterative scheme.

  14. Development of Momentum Conserving Monte Carlo Simulation Code for ECCD Study in Helical Plasmas

    Science.gov (United States)

    Murakami, S.; Hasegawa, S.; Moriya, Y.

    2015-03-01

    Parallel momentum conserving collision model is developed for GNET code, in which a linearized drift kinetic equation is solved in the five dimensional phase-space to study the electron cyclotron current drive (ECCD) in helical plasmas. In order to conserve the parallel momentum, we introduce a field particle collision term in addition to the test particle collision term. Two types of the field particle collision term are considered. One is the high speed limit model, where the momentum conserving term does not depend on the velocity of the background plasma and can be expressed in a simple form. The other is the velocity dependent model, which is derived from the Fokker-Planck collision term directly. In the velocity dependent model the field particle operator can be expressed using Legendre polynominals and, introducing the Rosenbluth potential, we derive the field particle term for each Legendre polynominals. In the GNET code, we introduce an iterative process to implement the momentum conserving collision operator. The high speed limit model is applied to the ECCD simulation of the heliotron-J plasma. The simulation results show a good conservation of the momentum with the iterative scheme.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ramsdell, James V.; Rishel, Jeremy P.

    2006-07-01

    This manual describes the atmospheric model and computer code for the Atmospheric Transport Module within SAC. The Atmospheric Transport Module, called RATCHET2, calculates the time-integrated air concentration and surface deposition of airborne contaminants to the soil. The RATCHET2 code is an adaptation of the Regional Atmospheric Transport Code for Hanford Emissions Tracking (RATCHET). The original RATCHET code was developed to perform the atmospheric transport for the Hanford Environmental Dose Reconstruction Project. Fundamentally, the two sets of codes are identical; no capabilities have been deleted from the original version of RATCHET. Most modifications are generally limited to revision of the run-specification file to streamline the simulation process for SAC.

  16. Optimization of GATE and PHITS Monte Carlo code parameters for uniform scanning proton beam based on simulation with FLUKA general-purpose code

    Energy Technology Data Exchange (ETDEWEB)

    Kurosu, Keita [Department of Medical Physics and Engineering, Osaka University Graduate School of Medicine, Suita, Osaka 565-0871 (Japan); Department of Radiation Oncology, Osaka University Graduate School of Medicine, Suita, Osaka 565-0871 (Japan); Takashina, Masaaki; Koizumi, Masahiko [Department of Medical Physics and Engineering, Osaka University Graduate School of Medicine, Suita, Osaka 565-0871 (Japan); Das, Indra J. [Department of Radiation Oncology, Indiana University School of Medicine, Indianapolis, IN 46202 (United States); Moskvin, Vadim P., E-mail: vadim.p.moskvin@gmail.com [Department of Radiation Oncology, Indiana University School of Medicine, Indianapolis, IN 46202 (United States)

    2014-10-01

    Although three general-purpose Monte Carlo (MC) simulation tools: Geant4, FLUKA and PHITS have been used extensively, differences in calculation results have been reported. The major causes are the implementation of the physical model, preset value of the ionization potential or definition of the maximum step size. In order to achieve artifact free MC simulation, an optimized parameters list for each simulation system is required. Several authors have already proposed the optimized lists, but those studies were performed with a simple system such as only a water phantom. Since particle beams have a transport, interaction and electromagnetic processes during beam delivery, establishment of an optimized parameters-list for whole beam delivery system is therefore of major importance. The purpose of this study was to determine the optimized parameters list for GATE and PHITS using proton treatment nozzle computational model. The simulation was performed with the broad scanning proton beam. The influences of the customizing parameters on the percentage depth dose (PDD) profile and the proton range were investigated by comparison with the result of FLUKA, and then the optimal parameters were determined. The PDD profile and the proton range obtained from our optimized parameters list showed different characteristics from the results obtained with simple system. This led to the conclusion that the physical model, particle transport mechanics and different geometry-based descriptions need accurate customization in planning computational experiments for artifact-free MC simulation.

  17. A graphics-card implementation of Monte-Carlo simulations for cosmic-ray transport

    Science.gov (United States)

    Tautz, R. C.

    2016-05-01

    A graphics card implementation of a test-particle simulation code is presented that is based on the CUDA extension of the C/C++ programming language. The original CPU version has been developed for the calculation of cosmic-ray diffusion coefficients in artificial Kolmogorov-type turbulence. In the new implementation, the magnetic turbulence generation, which is the most time-consuming part, is separated from the particle transport and is performed on a graphics card. In this article, the modification of the basic approach of integrating test particle trajectories to employ the SIMD (single instruction, multiple data) model is presented and verified. The efficiency of the new code is tested and several language-specific accelerating factors are discussed. For the example of isotropic magnetostatic turbulence, sample results are shown and a comparison to the results of the CPU implementation is performed.

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

    Energy Technology Data Exchange (ETDEWEB)

    VOOGD, J.A.

    1999-04-19

    An analysis of three software proposals is performed to recommend a computer code for immobilized low activity waste flow and transport modeling. The document uses criteria restablished in HNF-1839, ''Computer Code Selection Criteria for Flow and Transport Codes to be Used in Undisturbed Vadose Zone Calculation for TWRS Environmental Analyses'' as the basis for this analysis.

  19. Boltzmann equation analysis and Monte Carlo simulation of electron transport in N2-O2 streamer discharge

    NARCIS (Netherlands)

    Dujko, S.; Ebert, U.; White, R.D.; Petrović, Z.L.

    2010-01-01

    A comprehensive investigation of electron transport in N$_{2}$-O$_{2}$ mixtures has been carried out using a multi term theory for solving the Boltzmann equation and Monte Carlo simulation technique instead of conventional two-term theory often employed in plasma modeling community. We focus on the

  20. Validation of a GPU-based Monte Carlo code (gPMC) for proton radiation therapy: clinical cases study

    Science.gov (United States)

    Giantsoudi, Drosoula; Schuemann, Jan; Jia, Xun; Dowdell, Stephen; Jiang, Steve; Paganetti, Harald

    2015-03-01

    Monte Carlo (MC) methods are recognized as the gold-standard for dose calculation, however they have not replaced analytical methods up to now due to their lengthy calculation times. GPU-based applications allow MC dose calculations to be performed on time scales comparable to conventional analytical algorithms. This study focuses on validating our GPU-based MC code for proton dose calculation (gPMC) using an experimentally validated multi-purpose MC code (TOPAS) and compare their performance for clinical patient cases. Clinical cases from five treatment sites were selected covering the full range from very homogeneous patient geometries (liver) to patients with high geometrical complexity (air cavities and density heterogeneities in head-and-neck and lung patients) and from short beam range (breast) to large beam range (prostate). Both gPMC and TOPAS were used to calculate 3D dose distributions for all patients. Comparisons were performed based on target coverage indices (mean dose, V95, D98, D50, D02) and gamma index distributions. Dosimetric indices differed less than 2% between TOPAS and gPMC dose distributions for most cases. Gamma index analysis with 1%/1 mm criterion resulted in a passing rate of more than 94% of all patient voxels receiving more than 10% of the mean target dose, for all patients except for prostate cases. Although clinically insignificant, gPMC resulted in systematic underestimation of target dose for prostate cases by 1-2% compared to TOPAS. Correspondingly the gamma index analysis with 1%/1 mm criterion failed for most beams for this site, while for 2%/1 mm criterion passing rates of more than 94.6% of all patient voxels were observed. For the same initial number of simulated particles, calculation time for a single beam for a typical head and neck patient plan decreased from 4 CPU hours per million particles (2.8-2.9 GHz Intel X5600) for TOPAS to 2.4 s per million particles (NVIDIA TESLA C2075) for gPMC. Excellent agreement was

  1. Thermal neutron response of a boron-coated GEM detector via GEANT4 Monte Carlo code.

    Science.gov (United States)

    Jamil, M; Rhee, J T; Kim, H G; Ahmad, Farzana; Jeon, Y J

    2014-10-22

    In this work, we report the design configuration and the performance of the hybrid Gas Electron Multiplier (GEM) detector. In order to make the detector sensitive to thermal neutrons, the forward electrode of the GEM has been coated with the enriched boron-10 material, which works as a neutron converter. A total of 5×5cm(2) configuration of GEM has been used for thermal neutron studies. The response of the detector has been estimated via using GEANT4 MC code with two different physics lists. Using the QGSP_BIC_HP physics list, the neutron detection efficiency was determined to be about 3%, while with QGSP_BERT_HP physics list the efficiency was around 2.5%, at the incident thermal neutron energies of 25meV. The higher response of the detector proves that GEM-coated with boron converter improves the efficiency for thermal neutrons detection.

  2. Blind Decoding of Multiple Description Codes over OFDM Systems via Sequential Monte Carlo

    Directory of Open Access Journals (Sweden)

    Guo Dong

    2005-01-01

    Full Text Available We consider the problem of transmitting a continuous source through an OFDM system. Multiple description scalar quantization (MDSQ is applied to the source signal, resulting in two correlated source descriptions. The two descriptions are then OFDM modulated and transmitted through two parallel frequency-selective fading channels. At the receiver, a blind turbo receiver is developed for joint OFDM demodulation and MDSQ decoding. Transformation of the extrinsic information of the two descriptions are exchanged between each other to improve system performance. A blind soft-input soft-output OFDM detector is developed, which is based on the techniques of importance sampling and resampling. Such a detector is capable of exchanging the so-called extrinsic information with the other component in the above turbo receiver, and successively improving the overall receiver performance. Finally, we also treat channel-coded systems, and a novel blind turbo receiver is developed for joint demodulation, channel decoding, and MDSQ source decoding.

  3. BROMOCEA Code: An Improved Grand Canonical Monte Carlo/Brownian Dynamics Algorithm Including Explicit Atoms.

    Science.gov (United States)

    Solano, Carlos J F; Pothula, Karunakar R; Prajapati, Jigneshkumar D; De Biase, Pablo M; Noskov, Sergei Yu; Kleinekathöfer, Ulrich

    2016-05-10

    All-atom molecular dynamics simulations have a long history of applications studying ion and substrate permeation across biological and artificial pores. While offering unprecedented insights into the underpinning transport processes, MD simulations are limited in time-scales and ability to simulate physiological membrane potentials or asymmetric salt solutions and require substantial computational power. While several approaches to circumvent all of these limitations were developed, Brownian dynamics simulations remain an attractive option to the field. The main limitation, however, is an apparent lack of protein flexibility important for the accurate description of permeation events. In the present contribution, we report an extension of the Brownian dynamics scheme which includes conformational dynamics. To achieve this goal, the dynamics of amino-acid residues was incorporated into the many-body potential of mean force and into the Langevin equations of motion. The developed software solution, called BROMOCEA, was applied to ion transport through OmpC as a test case. Compared to fully atomistic simulations, the results show a clear improvement in the ratio of permeating anions and cations. The present tests strongly indicate that pore flexibility can enhance permeation properties which will become even more important in future applications to substrate translocation.

  4. STUDI PEMODELAN DAN PERHITUNGAN TRANSPORT MONTE CARLO DALAM TERAS HTR PEBBLE BED

    Directory of Open Access Journals (Sweden)

    Zuhair .

    2013-01-01

    Full Text Available Konsep sistem energi VHTR baik yang berbahan bakar pebble (VHTR pebble bed maupun blok prismatik (VHTR prismatik menarik perhatian fisikawan reaktor nuklir. Salah satu kelebihan teknologi bahan bakar bola adalah menawarkan terobosan teknologi pengisian bahan bakar tanpa harus menghentikan produksi listrik. Selain itu, partikel bahan bakar pebble dengan kernel uranium oksida (UO2 atau uranium oksikarbida (UCO yang dibalut TRISO dan pelapisan silikon karbida (SiC dianggap sebagai opsi utama dengan pertimbangan performa tinggi pada burn-up bahan bakar dan temperatur tinggi. Makalah ini mendiskusikan pemodelan dan perhitungan transport Monte Carlo dalam teras HTR pebble bed. HTR pebble bed adalah reaktor berpendingin gas temperatur tinggi dan bermoderator grafit dengan kemampuan kogenerasi. Perhitungan dikerjakan dengan program MCNP5 pada temperatur 1200 K. Pustaka data nuklir energi kontinu ENDF/B-V dan ENDF/B-VI dimanfaatkan untuk melengkapi analisis. Hasil perhitungan secara keseluruhan menunjukkan konsistensi dengan nilai keff yang hampir sama untuk pustaka data nuklir yang digunakan. Pustaka ENDF/B-VI (66c selalu memproduksi keff lebih besar dibandingkan ENDF/B-V (50c maupun ENDF/B-VI (60c dengan bias kurang dari 0,25%. Kisi BCC memprediksi keff hampir selalu lebih kecil daripada kisi lainnya, khususnya FCC. Nilai keff kisi BCC lebih dekat dengan kisi FCC dengan bias kurang dari 0,19% sedangkan dengan kisi SH bias perhitungannya kurang dari 0,22%. Fraksi packing yang sedikit berbeda (BCC= 61%, SH= 60,459% tidak membuat bias perhitungan menjadi berbeda jauh. Estimasi keff ketiga model kisi menyimpulkan bahwa model BCC lebih bisa diadopsi dalam perhitungan HTR pebble bed dibandingkan model FCC dan SH. Verifikasi hasil estimasi ini perlu dilakukan dengan simulasi Monte Carlo atau bahkan program deterministik lainnya guna optimisasi perhitungan teras reaktor temperatur tinggi.   Kata-kunci: kernel, TRISO, bahan bakar pebble, HTR pebble bed

  5. Monte Carlo simulation using the PENELOPE code with an ant colony algorithm to study MOSFET detectors

    Energy Technology Data Exchange (ETDEWEB)

    Carvajal, M A; Palma, A J [Departamento de Electronica y Tecnologia de Computadores, Universidad de Granada, E-18071 Granada (Spain); Garcia-Pareja, S [Servicio de Radiofisica Hospitalaria, Hospital Regional Universitario ' Carlos Haya' , Avda Carlos Haya, s/n, E-29010 Malaga (Spain); Guirado, D [Servicio de RadiofIsica, Hospital Universitario ' San Cecilio' , Avda Dr Oloriz, 16, E-18012 Granada (Spain); Vilches, M [Servicio de Fisica y Proteccion Radiologica, Hospital Regional Universitario ' Virgen de las Nieves' , Avda Fuerzas Armadas, 2, E-18014 Granada (Spain); Anguiano, M; Lallena, A M [Departamento de Fisica Atomica, Molecular y Nuclear, Universidad de Granada, E-18071 Granada (Spain)], E-mail: carvajal@ugr.es, E-mail: garciapareja@gmail.com, E-mail: dguirado@ugr.es, E-mail: mvilches@ugr.es, E-mail: mangui@ugr.es, E-mail: ajpalma@ugr.es, E-mail: lallena@ugr.es

    2009-10-21

    In this work we have developed a simulation tool, based on the PENELOPE code, to study the response of MOSFET devices to irradiation with high-energy photons. The energy deposited in the extremely thin silicon dioxide layer has been calculated. To reduce the statistical uncertainties, an ant colony algorithm has been implemented to drive the application of splitting and Russian roulette as variance reduction techniques. In this way, the uncertainty has been reduced by a factor of {approx}5, while the efficiency is increased by a factor of above 20. As an application, we have studied the dependence of the response of the pMOS transistor 3N163, used as a dosimeter, with the incidence angle of the radiation for three common photons sources used in radiotherapy: a {sup 60}Co Theratron-780 and the 6 and 18 MV beams produced by a Mevatron KDS LINAC. Experimental and simulated results have been obtained for gantry angles of 0 deg., 15 deg., 30 deg., 45 deg., 60 deg. and 75 deg. The agreement obtained has permitted validation of the simulation tool. We have studied how to reduce the angular dependence of the MOSFET response by using an additional encapsulation made of brass in the case of the two LINAC qualities considered.

  6. Monte Carlo determination of the conversion coefficients Hp(3)/Ka in a right cylinder phantom with 'PENELOPE' code. Comparison with 'MCNP' simulations.

    Science.gov (United States)

    Daures, J; Gouriou, J; Bordy, J M

    2011-03-01

    This work has been performed within the frame of the European Union ORAMED project (Optimisation of RAdiation protection for MEDical staff). The main goal of the project is to improve standards of protection for medical staff for procedures resulting in potentially high exposures and to develop methodologies for better assessing and for reducing, exposures to medical staff. The Work Package WP2 is involved in the development of practical eye-lens dosimetry in interventional radiology. This study is complementary of the part of the ENEA report concerning the calculations with the MCNP-4C code of the conversion factors related to the operational quantity H(p)(3). In this study, a set of energy- and angular-dependent conversion coefficients (H(p)(3)/K(a)), in the newly proposed square cylindrical phantom made of ICRU tissue, have been calculated with the Monte-Carlo code PENELOPE and MCNP5. The H(p)(3) values have been determined in terms of absorbed dose, according to the definition of this quantity, and also with the kerma approximation as formerly reported in ICRU reports. At a low-photon energy (up to 1 MeV), the two results obtained with the two methods are consistent. Nevertheless, large differences are showed at a higher energy. This is mainly due to the lack of electronic equilibrium, especially for small angle incidences. The values of the conversion coefficients obtained with the MCNP-4C code published by ENEA quite agree with the kerma approximation calculations obtained with PENELOPE. We also performed the same calculations with the code MCNP5 with two types of tallies: F6 for kerma approximation and *F8 for estimating the absorbed dose that is, as known, due to secondary electrons. PENELOPE and MCNP5 results agree for the kerma approximation and for the absorbed dose calculation of H(p)(3) and prove that, for photon energies larger than 1 MeV, the transport of the secondary electrons has to be taken into account.

  7. Voxel2MCNP: software for handling voxel models for Monte Carlo radiation transport calculations.

    Science.gov (United States)

    Hegenbart, Lars; Pölz, Stefan; Benzler, Andreas; Urban, Manfred

    2012-02-01

    Voxel2MCNP is a program that sets up radiation protection scenarios with voxel models and generates corresponding input files for the Monte Carlo code MCNPX. Its technology is based on object-oriented programming, and the development is platform-independent. It has a user-friendly graphical interface including a two- and three-dimensional viewer. A row of equipment models is implemented in the program. Various voxel model file formats are supported. Applications include calculation of counting efficiency of in vivo measurement scenarios and calculation of dose coefficients for internal and external radiation scenarios. Moreover, anthropometric parameters of voxel models, for instance chest wall thickness, can be determined. Voxel2MCNP offers several methods for voxel model manipulations including image registration techniques. The authors demonstrate the validity of the program results and provide references for previous successful implementations. The authors illustrate the reliability of calculated dose conversion factors and specific absorbed fractions. Voxel2MCNP is used on a regular basis to generate virtual radiation protection scenarios at Karlsruhe Institute of Technology while further improvements and developments are ongoing.

  8. Modification of PRETOR Code to Be Applied to Transport Simulation in Stellarators

    Energy Technology Data Exchange (ETDEWEB)

    Fontanet, J.; Castejon, F.; Dies, J.; Fontdecaba, J.; Alejaldre, C.

    2001-07-01

    The 1.5 D transport code PRETOR, that has been previously used to simulate tokamak plasmas, has been modified to perform transport analysis in stellarator geometry. The main modifications that have been introduced in the code are related with the magnetic equilibrium and with the modelling of energy and particle transport. Therefore, PRETOR- Stellarator version has been achieved and the code is suitable to perform simulations on stellarator plasmas. As an example, PRETOR- Stellarator has been used in the transport analysis of several Heliac Flexible TJ-II shots, and the results are compared with those obtained using PROCTR code. These results are also compared with the obtained using the tokamak version of PRETOR to show the importance of the introduced changes. (Author) 18 refs.

  9. Spallation integral experiment analysis by high energy nucleon-meson transport code

    Energy Technology Data Exchange (ETDEWEB)

    Takada, Hiroshi; Meigo, Shin-ichiro; Sasa, Toshinobu; Fukahori, Tokio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yoshizawa, Nobuaki; Furihata, Shiori; Belyakov-Bodin, V.I.; Krupny, G.I.; Titarenko, Y.E.

    1997-03-01

    Reaction rate distributions were measured with various activation detectors on the cylindrical surface of the thick tungsten target of 20 cm in diameter and 60 cm in length bombarded with the 0.895 and 1.21 GeV protons. The experimental results were analyzed with the Monte Carlo simulation code systems of NMTC/JAERI-MCNP-4A, LAHET and HERMES. It is confirmed that those code systems can represent the reaction rate distributions with the C/E ratio of 0.6 to 1.4 at the positions up to 30 cm from beam incident surface. (author)

  10. Investigation of Dosimetric Parameters of $^{192}$Ir MicroSelectron v2 HDR Brachytherapy Source Using EGSnrc Monte Carlo Code

    CERN Document Server

    Naeem, Hamza; Zheng, Huaqing; Cao, Ruifen; Pei, Xi; Hu, Liqin; Wu, Yican

    2016-01-01

    The $^{192}$Ir sources are widely used for high dose rate (HDR) brachytherapy treatments. The aim of this study is to simulate $^{192}$Ir MicroSelectron v2 HDR brachytherapy source and calculate the air kerma strength, dose rate constant, radial dose function and anisotropy function established in the updated AAPM Task Group 43 protocol. The EGSnrc Monte Carlo (MC) code package is used to calculate these dosimetric parameters, including dose contribution from secondary electron source and also contribution of bremsstrahlung photons to air kerma strength. The Air kerma strength, dose rate constant and radial dose function while anisotropy functions for the distance greater than 0.5 cm away from the source center are in good agreement with previous published studies. Obtained value from MC simulation for air kerma strength is $9.762\\times 10^{-8} \\textrm{UBq}^{-1}$and dose rate constant is $1.108\\pm 0.13\\%\\textrm{cGyh}^{-1} \\textrm{U}^{-1}$.

  11. EleCa: a Monte Carlo code for the propagation of extragalactic photons at ultra-high energy

    CERN Document Server

    Settimo, Mariangela

    2013-01-01

    Ultra high energy photons play an important role as an independent probe of the photo-pion production mechanism by UHE cosmic rays. Their observation, or non-observation, may constrain astrophysical scenarios for the origin of UHECRs and help to understand the nature of the flux suppression observed by several experiments at energies above 10$^{19.5}$ eV. Whereas the interaction length of UHE photons above 10$^{17}$ eV is only of a few hundred kpc up to tenths of Mpc, photons can interact with the extragalactic background radiation leading to the development of electromagnetic cascades which affect the fluxes of photons observed at Earth. The interpretation of the current experimental results rely on the simulations of the UHE photon propagation. In this contribution, we present the novel Monte Carlo code "EleCa" to simulate the \\emph{Ele}ctromagnetic \\emph{Ca}scading initiated by high-energy photons and electrons. The distance within which we expect to observe UHE photons is discussed and the flux of GZK pho...

  12. The Wigner Monte-Carlo method for nanoelectronic devices a particle description of quantum transport and decoherence

    CERN Document Server

    Querlioz, Damien

    2013-01-01

    This book gives an overview of the quantum transport approaches for nanodevices and focuses on the Wigner formalism. It details the implementation of a particle-based Monte Carlo solution of the Wigner transport equation and how the technique is applied to typical devices exhibiting quantum phenomena, such as the resonant tunnelling diode, the ultra-short silicon MOSFET and the carbon nanotube transistor. In the final part, decoherence theory is used to explain the emergence of the semi-classical transport in nanodevices.

  13. Study of cold neutron sources: Implementation and validation of a complete computation scheme for research reactor using Monte Carlo codes TRIPOLI-4.4 and McStas

    Energy Technology Data Exchange (ETDEWEB)

    Campioni, Guillaume; Mounier, Claude [Commissariat a l' Energie Atomique, CEA, 31-33, rue de la Federation, 75752 Paris cedex (France)

    2006-07-01

    The main goal of the thesis about studies of cold neutrons sources (CNS) in research reactors was to create a complete set of tools to design efficiently CNS. The work raises the problem to run accurate simulations of experimental devices inside reactor reflector valid for parametric studies. On one hand, deterministic codes have reasonable computation times but introduce problems for geometrical description. On the other hand, Monte Carlo codes give the possibility to compute on precise geometry, but need computation times so important that parametric studies are impossible. To decrease this computation time, several developments were made in the Monte Carlo code TRIPOLI-4.4. An uncoupling technique is used to isolate a study zone in the complete reactor geometry. By recording boundary conditions (incoming flux), further simulations can be launched for parametric studies with a computation time reduced by a factor 60 (case of the cold neutron source of the Orphee reactor). The short response time allows to lead parametric studies using Monte Carlo code. Moreover, using biasing methods, the flux can be recorded on the surface of neutrons guides entries (low solid angle) with a further gain of running time. Finally, the implementation of a coupling module between TRIPOLI- 4.4 and the Monte Carlo code McStas for research in condensed matter field gives the possibility to obtain fluxes after transmission through neutrons guides, thus to have the neutron flux received by samples studied by scientists of condensed matter. This set of developments, involving TRIPOLI-4.4 and McStas, represent a complete computation scheme for research reactors: from nuclear core, where neutrons are created, to the exit of neutrons guides, on samples of matter. This complete calculation scheme is tested against ILL4 measurements of flux in cold neutron guides. (authors)

  14. Monte Carlo modeling of photon transport in buried bone tissue layer for quantitative Raman spectroscopy

    Science.gov (United States)

    Wilson, Robert H.; Dooley, Kathryn A.; Morris, Michael D.; Mycek, Mary-Ann

    2009-02-01

    Light-scattering spectroscopy has the potential to provide information about bone composition via a fiber-optic probe placed on the skin. In order to design efficient probes, one must understand the effect of all tissue layers on photon transport. To quantitatively understand the effect of overlying tissue layers on the detected bone Raman signal, a layered Monte Carlo model was modified for Raman scattering. The model incorporated the absorption and scattering properties of three overlying tissue layers (dermis, subdermis, muscle), as well as the underlying bone tissue. The attenuation of the collected bone Raman signal, predominantly due to elastic light scattering in the overlying tissue layers, affected the carbonate/phosphate (C/P) ratio by increasing the standard deviation of the computational result. Furthermore, the mean C/P ratio varied when the relative thicknesses of the layers were varied and the elastic scattering coefficient at the Raman scattering wavelength of carbonate was modeled to be different from that at the Raman scattering wavelength of phosphate. These results represent the first portion of a computational study designed to predict optimal probe geometry and help to analyze detected signal for Raman scattering experiments involving bone.

  15. Comparison of some popular Monte Carlo solution for proton transportation within pCT problem

    Energy Technology Data Exchange (ETDEWEB)

    Evseev, Ivan; Assis, Joaquim T. de; Yevseyeva, Olga [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Inst. Politecnico], E-mail: evseev@iprj.uerj.br, E-mail: joaquim@iprj.uerj.br, E-mail: yevseyeva@iprj.uerj.br; Lopes, Ricardo T.; Cardoso, Jose J.B.; Silva, Ademir X. da [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE). Lab. de Instrumentacao Nuclear], E-mail: ricardo@lin.ufrj.br, E-mail: jjbrum@oi.com.br, E-mail: ademir@con.ufrj.br; Vinagre Filho, Ubirajara M. [Instituto de Engenharia Nuclear IEN/CNEN-RJ, Rio de Janeiro, RJ (Brazil)], E-mail: bira@ien.gov.br; Hormaza, Joel M. [UNESP, Botucatu, SP (Brazil). Inst. de Biociencias], E-mail: jmesa@ibb.unesp.br; Schelin, Hugo R.; Paschuk, Sergei A.; Setti, Joao A.P.; Milhoretto, Edney [Universidade Tecnologica Federal do Parana, Curitiba, PR (Brazil)], E-mail: schelin@cpgei.cefetpr.br, E-mail: sergei@utfpr.edu.br, E-mail: jsetti@gmail.com, E-mail: edneymilhoretto@yahoo.com

    2007-07-01

    The proton transport in matter is described by the Boltzmann kinetic equation for the proton flux density. This equation, however, does not have a general analytical solution. Some approximate analytical solutions have been developed within a number of significant simplifications. Alternatively, the Monte Carlo simulations are widely used. Current work is devoted to the discussion of the proton energy spectra obtained by simulation with SRIM2006, GEANT4 and MCNPX packages. The simulations have been performed considering some further applications of the obtained results in computed tomography with proton beam (pCT). Thus the initial and outgoing proton energies (3 / 300 MeV) as well as the thickness of irradiated target (water and aluminum phantoms within 90% of the full range for a given proton beam energy) were considered in the interval of values typical for pCT applications. One from the most interesting results of this comparison is that while the MCNPX spectra are in a good agreement with analytical description within Fokker-Plank approximation and the GEANT4 simulated spectra are slightly shifted from them the SRIM2006 simulations predict a notably higher mean energy loss for protons. (author)

  16. Core-scale solute transport model selection using Monte Carlo analysis

    Science.gov (United States)

    Malama, Bwalya; Kuhlman, Kristopher L.; James, Scott C.

    2013-06-01

    Model applicability to core-scale solute transport is evaluated using breakthrough data from column experiments conducted with conservative tracers tritium (3H) and sodium-22 (22Na ), and the retarding solute uranium-232 (232U). The three models considered are single-porosity, double-porosity with single-rate mobile-immobile mass-exchange, and the multirate model, which is a deterministic model that admits the statistics of a random mobile-immobile mass-exchange rate coefficient. The experiments were conducted on intact Culebra Dolomite core samples. Previously, data were analyzed using single-porosity and double-porosity models although the Culebra Dolomite is known to possess multiple types and scales of porosity, and to exhibit multirate mobile-immobile-domain mass transfer characteristics at field scale. The data are reanalyzed here and null-space Monte Carlo analysis is used to facilitate objective model selection. Prediction (or residual) bias is adopted as a measure of the model structural error. The analysis clearly shows single-porosity and double-porosity models are structurally deficient, yielding late-time residual bias that grows with time. On the other hand, the multirate model yields unbiased predictions consistent with the late-time -5/2 slope diagnostic of multirate mass transfer. The analysis indicates the multirate model is better suited to describing core-scale solute breakthrough in the Culebra Dolomite than the other two models.

  17. Detailed Monte Carlo Simulation of electron transport and electron energy loss spectra.

    Science.gov (United States)

    Attarian Shandiz, M; Salvat, F; Gauvin, R

    2016-11-01

    A computer program for detailed Monte Carlo simulation of the transport of electrons with kinetic energies in the range between about 0.1 and about 500 keV in bulk materials and in thin solid films is presented. Elastic scattering is described from differential cross sections calculated by the relativistic (Dirac) partial-wave expansion method with different models of the scattering potential. Inelastic interactions are simulated from an optical-data model based on an empirical optical oscillator strength that combines optical functions of the solid with atomic photoelectric data. The generalized oscillator strength is built from the adopted optical oscillator strength by using an extension algorithm derived from Lindhard's dielectric function for a free-electron gas. It is shown that simulated backscattering fractions of electron beams from bulk (semi-infinite) specimens are in good agreement with experimental data for beam energies from 0.1 keV up to about 100 keV. Simulations also yield transmitted and backscattered fractions of electron beams on thin solid films that agree closely with measurements for different film thicknesses and incidence angles. Simulated most probable deflection angles and depth-dose distributions also agree satisfactorily with measurements. Finally, electron energy loss spectra of several elemental solids are simulated and the effects of the beam energy and the foil thickness on the signal to background and signal to noise ratios are investigated. SCANNING 38:475-491, 2016. © 2015 Wiley Periodicals, Inc.

  18. Environmental, Transient, Three-Dimensional, Hydrothermal, Mass Transport Code - FLESCOT

    Energy Technology Data Exchange (ETDEWEB)

    Onishi, Yasuo [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bao, Jie [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Glass, Kevin A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Eyler, L. L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Okumura, Masahiko [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-03-28

    The purpose of the project was to modify and apply the transient, three-dimensional FLESCOT code to be able to effectively simulate cesium behavior in Fukushima lakes/dam reservoirs, river mouths, and coastal areas. The ultimate objective of the FLESCOT simulation is to predict future changes of cesium accumulation in Fukushima area reservoirs and costal water. These evaluation results will assist ongoing and future environmental remediation activities and policies in a systematic and comprehensive manner.

  19. LEADS-DC: A computer code for intense dc beam nonlinear transport simulation

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    An intense dc beam nonlinear transport code has been developed. The code is written in Visual FORTRAN 6.6 and has ~13000 lines. The particle distribution in the transverse cross section is uniform or Gaussian. The space charge forces are calculated by the PIC (particle in cell) scheme, and the effects of the applied fields on the particle motion are calculated with the Lie algebraic method through the third order approximation. Obviously,the solutions to the equations of particle motion are self-consistent. The results obtained from the theoretical analysis have been put in the computer code. Many optical beam elements are contained in the code. So, the code can simulate the intense dc particle motions in the beam transport lines, high voltage dc accelerators and ion implanters.

  20. The TOUGH codes - a family of simulation tools for multiphase flowand transport processes in permeable media

    Energy Technology Data Exchange (ETDEWEB)

    Pruess, Karsten

    2003-08-08

    Numerical simulation has become a widely practiced andaccepted technique for studying flow and transport processes in thevadose zone and other subsurface flow systems. This article discusses asuite of codes, developed primarily at Lawrence Berkeley NationalLaboratory (LBNL), with the capability to model multiphase flows withphase change. We summarize history and goals in the development of theTOUGH codes, and present the governing equations for multiphase,multicomponent flow. Special emphasis is given to space discretization bymeans of integral finite differences (IFD). Issues of code implementationand architecture are addressed, as well as code applications,maintenance, and future developments.

  1. A CUMULATIVE MIGRATION METHOD FOR COMPUTING RIGOROUS TRANSPORT CROSS SECTIONS AND DIFFUSION COEFFICIENTS FOR LWR LATTICES WITH MONTE CARLO

    Energy Technology Data Exchange (ETDEWEB)

    Zhaoyuan Liu; Kord Smith; Benoit Forget; Javier Ortensi

    2016-05-01

    A new method for computing homogenized assembly neutron transport cross sections and dif- fusion coefficients that is both rigorous and computationally efficient is proposed in this paper. In the limit of a homogeneous hydrogen slab, the new method is equivalent to the long-used, and only-recently-published CASMO transport method. The rigorous method is used to demonstrate the sources of inaccuracy in the commonly applied “out-scatter” transport correction. It is also demonstrated that the newly developed method is directly applicable to lattice calculations per- formed by Monte Carlo and is capable of computing rigorous homogenized transport cross sections for arbitrarily heterogeneous lattices. Comparisons of several common transport cross section ap- proximations are presented for a simple problem of infinite medium hydrogen. The new method has also been applied in computing 2-group diffusion data for an actual PWR lattice from BEAVRS benchmark.

  2. CONDENSED MONTE-CARLO SIMULATIONS FOR THE DESCRIPTION OF LIGHT TRANSPORT

    NARCIS (Netherlands)

    GRAAFF, R; KOELINK, MH; DEMUL, FFM; ZIJLSTRA, WG; DASSEL, ACM; AARNOUDSE, JG

    1993-01-01

    A novel method, condensed Monte Carlo simulation, is presented that applies the results of a single Monte Carlo simulation for a given albedo mu(s)/(mu(a) + mu(s)) to obtaining results for other albedos; mu(s) and mu(a) are the scattering and absorption coefficients, respectively. The method require

  3. Numerical model for two-dimensional hydrodynamics and energy transport. [VECTRA code

    Energy Technology Data Exchange (ETDEWEB)

    Trent, D.S.

    1973-06-01

    The theoretical basis and computational procedure of the VECTRA computer program are presented. VECTRA (Vorticity-Energy Code for TRansport Analysis) is designed for applying numerical simulation to a broad range of intake/discharge flows in conjunction with power plant hydrological evaluation. The code computational procedure is based on finite-difference approximation of the vorticity-stream function partial differential equations which govern steady flow momentum transport of two-dimensional, incompressible, viscous fluids in conjunction with the transport of heat and other constituents.

  4. A study of the earth radiation budget using a 3D Monte-Carlo radiative transer code

    Science.gov (United States)

    Okata, M.; Nakajima, T.; Sato, Y.; Inoue, T.; Donovan, D. P.

    2013-12-01

    The purpose of this study is to evaluate the earth's radiation budget when data are available from satellite-borne active sensors, i.e. cloud profiling radar (CPR) and lidar, and a multi-spectral imager (MSI) in the project of the Earth Explorer/EarthCARE mission. For this purpose, we first developed forward and backward 3D Monte Carlo radiative transfer codes that can treat a broadband solar flux calculation including thermal infrared emission calculation by k-distribution parameters of Sekiguchi and Nakajima (2008). In order to construct the 3D cloud field, we tried the following three methods: 1) stochastic cloud generated by randomized optical thickness each layer distribution and regularly-distributed tilted clouds, 2) numerical simulations by a non-hydrostatic model with bin cloud microphysics model and 3) Minimum cloud Information Deviation Profiling Method (MIDPM) as explained later. As for the method-2 (numerical modeling method), we employed numerical simulation results of Californian summer stratus clouds simulated by a non-hydrostatic atmospheric model with a bin-type cloud microphysics model based on the JMA NHM model (Iguchi et al., 2008; Sato et al., 2009, 2012) with horizontal (vertical) grid spacing of 100m (20m) and 300m (20m) in a domain of 30km (x), 30km (y), 1.5km (z) and with a horizontally periodic lateral boundary condition. Two different cell systems were simulated depending on the cloud condensation nuclei (CCN) concentration. In the case of horizontal resolution of 100m, regionally averaged cloud optical thickness, , and standard deviation of COT, were 3.0 and 4.3 for pristine case and 8.5 and 7.4 for polluted case, respectively. In the MIDPM method, we first construct a library of pair of observed vertical profiles from active sensors and collocated imager products at the nadir footprint, i.e. spectral imager radiances, cloud optical thickness (COT), effective particle radius (RE) and cloud top temperature (Tc). We then select a best

  5. Modeling and commissioning of a Clinac 600 CD by Monte Carlo method using the BEAMnrc and DOSXYZnrc codes

    Energy Technology Data Exchange (ETDEWEB)

    Junior, Reginaldo G., E-mail: reginaldo.junior@ifmg.edu.br [Instituto Federal de Minas Gerais (IFMG), Formiga, MG (Brazil). Departamento de Engenharia Eletrica; Oliveira, Arno H. de; Sousa, Romulo V., E-mail: arnoheeren@gmail.com, E-mail: romuloverdolin@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Mourao, Arnaldo P., E-mail: apratabhz@gmail.com [Centro Federal de Educacao Tecnologica de Minas Gerais, Belo Horizonte, MG (Brazil)

    2015-07-01

    This paper reports the modeling of a linear accelerator Clinac 600 CD with BEAMnrc application, derived from EGSnrc radiation transport code, indicating relevant details of modeling that traditionally involve difficulties imposed on the process. This accelerator was commissioned by the confrontation of experimental dosimetric data with the computer data obtained by DOSXYZnrc application. The information compared in dosimetry process were: field profiles and dose percentage curves obtained in a water phantom with cubic edge of 30 cm. In all comparisons made, the computational data showed satisfactory precision and discrepancies with the experimental data did not exceed 3%, proving the electiveness of the model. Both the accelerator model and the computational dosimetry methodology, revealed the need for adjustments that probably will allow obtaining more accurate data than those obtained in the simulations presented here. These adjustments are mainly associated to improve the resolution of the eld profiles, the voxelization in phantom and optimization of computing time. (author)

  6. Application of a non-steady-state orbit-following Monte-Carlo code to neutron modeling in the MAST spherical tokamak

    Science.gov (United States)

    Tani, K.; Shinohara, K.; Oikawa, T.; Tsutsui, H.; McClements, K. G.; Akers, R. J.; Liu, Y. Q.; Suzuki, M.; Ide, S.; Kusama, Y.; Tsuji-Iio, S.

    2016-11-01

    As part of the verification and validation of a newly developed non-steady-state orbit-following Monte-Carlo code, application studies of time dependent neutron rates have been made for a specific shot in the Mega Amp Spherical Tokamak (MAST) using 3D fields representing vacuum resonant magnetic perturbations (RMPs) and toroidal field (TF) ripples. The time evolution of density, temperature and rotation rate in the application of the code to MAST are taken directly from experiment. The calculation results approximately agree with the experimental data. It is also found that a full orbit-following scheme is essential to reproduce the neutron rates in MAST.

  7. Absorbed dose estimations of 131I for critical organs using the GEANT4 Monte Carlo simulation code

    Institute of Scientific and Technical Information of China (English)

    Ziaur Rahman; Shakeel ur Rehman; Waheed Arshed; Nasir M Mirza; Abdul Rashid; Jahan Zeb

    2012-01-01

    The aim of this study is to compare the absorbed doses of critical organs of 131I using the MIRD (Medical Internal Radiation Dose) with the corresponding predictions made by GEANT4 simulations.S-values (mean absorbed dose rate per unit activity) and energy deposition per decay for critical organs of 131I for various ages,using standard cylindrical phantom comprising water and ICRP soft-tissue material,have also been estimated.In this study the effect of volume reduction of thyroid,during radiation therapy,on the calculation of absorbed dose is also being estimated using GEANT4.Photon specific energy deposition in the other organs of the neck,due to 131I decay in the thyroid organ,has also been estimated.The maximum relative difference of MIRD with the GEANT4 simulated results is 5.64% for an adult's critical organs of 131I.Excellent agreement was found between the results of water and ICRP soft tissue using the cylindrical model.S-values are tabulated for critical organs of 131I,using 1,5,10,15 and 18 years (adults) individuals.S-values for a cylindrical thyroid of different sizes,having 3.07% relative differences of GEANT4 with Siegel & Stabin results.Comparison of the experimentally measured values at 0.5 and 1 m away from neck of the ionization chamber with GEANT4 based Monte Carlo simulations results show good agreement.This study shows that GEANT4 code is an important tool for the internal dosimetry calculations.

  8. State-of-the-art Monte Carlo 1988

    Energy Technology Data Exchange (ETDEWEB)

    Soran, P.D.

    1988-06-28

    Particle transport calculations in highly dimensional and physically complex geometries, such as detector calibration, radiation shielding, space reactors, and oil-well logging, generally require Monte Carlo transport techniques. Monte Carlo particle transport can be performed on a variety of computers ranging from APOLLOs to VAXs. Some of the hardware and software developments, which now permit Monte Carlo methods to be routinely used, are reviewed in this paper. The development of inexpensive, large, fast computer memory, coupled with fast central processing units, permits Monte Carlo calculations to be performed on workstations, minicomputers, and supercomputers. The Monte Carlo renaissance is further aided by innovations in computer architecture and software development. Advances in vectorization and parallelization architecture have resulted in the development of new algorithms which have greatly reduced processing times. Finally, the renewed interest in Monte Carlo has spawned new variance reduction techniques which are being implemented in large computer codes. 45 refs.

  9. Transport Corrections in Nodal Diffusion Codes for HTR Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Abderrafi M. Ougouag; Frederick N. Gleicher

    2010-08-01

    The cores and reflectors of High Temperature Reactors (HTRs) of the Next Generation Nuclear Plant (NGNP) type are dominantly diffusive media from the point of view of behavior of the neutrons and their migration between the various structures of the reactor. This means that neutron diffusion theory is sufficient for modeling most features of such reactors and transport theory may not be needed for most applications. Of course, the above statement assumes the availability of homogenized diffusion theory data. The statement is true for most situations but not all. Two features of NGNP-type HTRs require that the diffusion theory-based solution be corrected for local transport effects. These two cases are the treatment of burnable poisons (BP) in the case of the prismatic block reactors and, for both pebble bed reactor (PBR) and prismatic block reactor (PMR) designs, that of control rods (CR) embedded in non-multiplying regions near the interface between fueled zones and said non-multiplying zones. The need for transport correction arises because diffusion theory-based solutions appear not to provide sufficient fidelity in these situations.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-02-01

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

  11. 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.

  12. MCNP-X Monte Carlo Code Application for Mass Attenuation Coefficients of Concrete at Different Energies by Modeling 3 × 3 Inch NaI(Tl Detector and Comparison with XCOM and Monte Carlo Data

    Directory of Open Access Journals (Sweden)

    Huseyin Ozan Tekin

    2016-01-01

    Full Text Available Gamma-ray measurements in various research fields require efficient detectors. One of these research fields is mass attenuation coefficients of different materials. Apart from experimental studies, the Monte Carlo (MC method has become one of the most popular tools in detector studies. An NaI(Tl detector has been modeled, and, for a validation study of the modeled NaI(Tl detector, the absolute efficiency of 3 × 3 inch cylindrical NaI(Tl detector has been calculated by using the general purpose Monte Carlo code MCNP-X (version 2.4.0 and compared with previous studies in literature in the range of 661–2620 keV. In the present work, the applicability of MCNP-X Monte Carlo code for mass attenuation of concrete sample material as building material at photon energies 59.5 keV, 80 keV, 356 keV, 661.6 keV, 1173.2 keV, and 1332.5 keV has been tested by using validated NaI(Tl detector. The mass attenuation coefficients of concrete sample have been calculated. The calculated results agreed well with experimental and some other theoretical results. The results specify that this process can be followed to determine the data on the attenuation of gamma-rays with other required energies in other materials or in new complex materials. It can be concluded that data from Monte Carlo is a strong tool not only for efficiency studies but also for mass attenuation coefficients calculations.

  13. Verification and Validation of The Tritium Transport Code TMAP7

    Energy Technology Data Exchange (ETDEWEB)

    Glen R. Longhurst; James Ambrosek

    2004-09-01

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

  14. Conception and development of an adaptive energy mesher for multigroup library generation of the transport codes; Conception et developpement d'un mailleur energetique adaptatif pour la generation des bibliotheques multigroupes des codes de transport

    Energy Technology Data Exchange (ETDEWEB)

    Mosca, P.

    2009-12-15

    The deterministic transport codes solve the stationary Boltzmann equation in a discretized energy formalism called multigroup. The transformation of continuous data in a multigroup form is obtained by averaging the highly variable cross sections of the resonant isotopes with the solution of the self-shielding models and the remaining ones with the coarse energy spectrum of the reactor type. So far the error of such an approach could only be evaluated retrospectively. To remedy this, we studied in this thesis a set of methods to control a priori the accuracy and the cost of the multigroup transport computation. The energy mesh optimisation is achieved using a two step process: the creation of a reference mesh and its optimized condensation. In the first stage, by refining locally and globally the energy mesh, we seek, on a fine energy mesh with subgroup self-shielding, a solution equivalent to a reference solver (Monte Carlo or pointwise deterministic solver). In the second step, once fixed the number of groups, depending on the acceptable computational cost, and chosen the most appropriate self-shielding models to the reactor type, we look for the best bounds of the reference mesh minimizing reaction rate errors by the particle swarm optimization algorithm. This new approach allows us to define new meshes for fast reactors as accurate as the currently used ones, but with fewer groups. (author)

  15. Analysis of the neutrons dispersion in a semi-infinite medium based in transport theory and the Monte Carlo method; Analisis de la dispersion de neutrones en un medio semi-infinito en base a teoria de transporte y el metodo de Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Arreola V, G. [IPN, Escuela Superior de Fisica y Matematicas, Posgrado en Ciencias Fisicomatematicas, area en Ingenieria Nuclear, Unidad Profesional Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07730 Mexico D. F. (Mexico); Vazquez R, R.; Guzman A, J. R., E-mail: energia.arreola.uam@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2012-10-15

    In this work a comparative analysis of the results for the neutrons dispersion in a not multiplicative semi-infinite medium is presented. One of the frontiers of this medium is located in the origin of coordinates, where a neutrons source in beam form, i.e., {mu}{omicron}=1 is also. The neutrons dispersion is studied on the statistical method of Monte Carlo and through the unidimensional transport theory and for an energy group. The application of transport theory gives a semi-analytic solution for this problem while the statistical solution for the flow was obtained applying the MCNPX code. The dispersion in light water and heavy water was studied. A first remarkable result is that both methods locate the maximum of the neutrons distribution to less than two mean free trajectories of transport for heavy water, while for the light water is less than ten mean free trajectories of transport; the differences between both methods is major for the light water case. A second remarkable result is that the tendency of both distributions is similar in small mean free trajectories, while in big mean free trajectories the transport theory spreads to an asymptote value and the solution in base statistical method spreads to zero. The existence of a neutron current of low energy and toward the source is demonstrated, in contrary sense to the neutron current of high energy coming from the own source. (Author)

  16. Development of Burnup Calculation Function in Reactor Monte Carlo Code RMC%堆用蒙卡程序燃耗计算功能开发

    Institute of Scientific and Technical Information of China (English)

    佘顶; 王侃; 余纲林

    2012-01-01

    This paper presents the burnup calculation capability of RMC, which is a new Monte Carlo (MC) neutron transport code developed by Reactor Engineering Analysis Laboratory (REAL) in Tsinghua university of China. Unlike most of existing MC depletion codes which explicitly couple the depletion module, RMC incorporates ORIGEN 2.1 in an implicit way. Different burn step strategies, including the middle-of-step approximation and the predictor-corrector method, are adopted by RMC to assure the accuracy under large burnup step size. RMC employs a spectrum-based method of tallying one-group cross section, which can considerably saves computational time with negligible accuracy loss. According to the validation results of benchmarks and examples, it is proved that the burnup function of RMC performs quite well in accuracy and efficiency.%堆用蒙卡程序(RMC)是由清华大学工程物理系REAL实验室自主开发的用于反应堆物理分析的中子输运蒙卡程序,本文主要介绍其燃耗计算功能的开发与验证.RMC的燃耗计算功能具有的特点:内部耦合ORIGEN,相比于外耦合方式,更加灵活和高效;使用基于能谱的单群截面统计方法,可在保证精度的前提下,显著提高计算效率;采取预估修正和中点近似等多种燃耗步策略,减小大燃耗步长时的计算误差.通过计算压水堆栅元、沸水堆组件、快堆等一系列基准题和算例,验证了RMC燃耗计算的正确性和速度优势.

  17. Carrier transport in dichromatic color-coded semipolar (2021) and (2021) III-N LEDs

    Science.gov (United States)

    Kisin, Mikhail V.; Huang, Chih-Li; El-Ghoroury, Hussein S.

    2014-03-01

    Simulation of III-nitride color-coded multiple quantum well (MQW) LED structures was performed using as an experimental benchmark dichromatic semipolar LEDs grown in Ga-polar and N-polar crystallographic orientations (Y. Kawaguchi et.al, APL 100, 231110, 2012). Different QW depths in the color-coded LEDs and opposite interface polarization charges in Ga-polar and N-polar structures provide different conditions for carrier transport across the LED active regions. Combination of several effects was crucial for adequate reproduction of the emission spectra experimentally observed in color-coded structures with violet-aquamarine and aquamarine-violet active region layouts. A standard drift-diffusion transport model wascompleted with rate equations for nonequilibrium QW populations and several high-energy transport features, including the effects of QW carrier overshoot and Auger-assisted QW depopulation. COMSOL-based Optoelectronic Device Modeling Software (ODMS) developed at Ostendo Technologies Inc. was utilized for device simulation.

  18. Simulation of clinical X-ray tube using the Monte Carlo Method - PENELOPE code; Simulacao de um tubo de raios X clinico atraves do Metodo de Monte Carlo usando codigo PENELOPE

    Energy Technology Data Exchange (ETDEWEB)

    Albuquerque, M.A.G.; David, M.G.; Almeida, C.E. de; Magalhaes, L.A.G., E-mail: malbuqueque@hotmail.com [Universidade do Estado do Rio de Janeiro (UERJ), Rio de Janeiro, RJ (Brazil). Lab. de Ciencias Radiologicas; Bernal, M. [Universidade Estadual de Campinas (UNICAMP), SP (Brazil); Braz, D. [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil)

    2015-07-01

    Breast cancer is the most common type of cancer among women. The main strategy to increase the long-term survival of patients with this disease is the early detection of the tumor, and mammography is the most appropriate method for this purpose. Despite the reduction of cancer deaths, there is a big concern about the damage caused by the ionizing radiation to the breast tissue. To evaluate these measures it was modeled a mammography equipment, and obtained the depth spectra using the Monte Carlo method - PENELOPE code. The average energies of the spectra in depth and the half value layer of the mammography output spectrum. (author)

  19. Dose calculations for a simplified Mammosite system with the Monte Carlo Penelope and MCNPX simulation codes; Calculos de dosis para un sistema Mammosite simplificado con los codigos de simulacion Monte Carlo PENELOPE y MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Rojas C, E.L.; Varon T, C.F.; Pedraza N, R. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: elrc@nuclear.inin.mx

    2007-07-01

    The treatment of the breast cancer at early stages is of vital importance. For that, most of the investigations are dedicated to the early detection of the suffering and their treatment. As investigation consequence and clinical practice, in 2002 it was developed in U.S.A. an irradiation system of high dose rate known as Mammosite. In this work we carry out dose calculations for a simplified Mammosite system with the Monte Carlo Penelope simulation code and MCNPX, varying the concentration of the contrast material that it is used in the one. (Author)

  20. A Modular Computer Code for Simulating Reactive Multi-Species Transport in 3-Dimensional Groundwater Systems

    Energy Technology Data Exchange (ETDEWEB)

    TP Clement

    1999-06-24

    RT3DV1 (Reactive Transport in 3-Dimensions) is computer code that solves the coupled partial differential equations that describe reactive-flow and transport of multiple mobile and/or immobile species in three-dimensional saturated groundwater systems. RT3D is a generalized multi-species version of the US Environmental Protection Agency (EPA) transport code, MT3D (Zheng, 1990). The current version of RT3D uses the advection and dispersion solvers from the DOD-1.5 (1997) version of MT3D. As with MT3D, RT3D also requires the groundwater flow code MODFLOW for computing spatial and temporal variations in groundwater head distribution. The RT3D code was originally developed to support the contaminant transport modeling efforts at natural attenuation demonstration sites. As a research tool, RT3D has also been used to model several laboratory and pilot-scale active bioremediation experiments. The performance of RT3D has been validated by comparing the code results against various numerical and analytical solutions. The code is currently being used to model field-scale natural attenuation at multiple sites. The RT3D code is unique in that it includes an implicit reaction solver that makes the code sufficiently flexible for simulating various types of chemical and microbial reaction kinetics. RT3D V1.0 supports seven pre-programmed reaction modules that can be used to simulate different types of reactive contaminants including benzene-toluene-xylene mixtures (BTEX), and chlorinated solvents such as tetrachloroethene (PCE) and trichloroethene (TCE). In addition, RT3D has a user-defined reaction option that can be used to simulate any other types of user-specified reactive transport systems. This report describes the mathematical details of the RT3D computer code and its input/output data structure. It is assumed that the user is familiar with the basics of groundwater flow and contaminant transport mechanics. In addition, RT3D users are expected to have some experience in

  1. Research on GPU Acceleration for Monte Carlo Criticality Calculation

    Science.gov (United States)

    Xu, Qi; Yu, Ganglin; Wang, Kan

    2014-06-01

    The Monte Carlo neutron transport method can be naturally parallelized by multi-core architectures due to the dependency between particles during the simulation. The GPU+CPU heterogeneous parallel mode has become an increasingly popular way of parallelism in the field of scientific supercomputing. Thus, this work focuses on the GPU acceleration method for the Monte Carlo criticality simulation, as well as the computational efficiency that GPUs can bring. The "neutron transport step" is introduced to increase the GPU thread occupancy. In order to test the sensitivity of the MC code's complexity, a 1D one-group code and a 3D multi-group general purpose code are respectively transplanted to GPUs, and the acceleration effects are compared. The result of numerical experiments shows considerable acceleration effect of the "neutron transport step" strategy. However, the performance comparison between the 1D code and the 3D code indicates the poor scalability of MC codes on GPUs.

  2. Some Examples of the Application and Validation of the NUFT Subsurface Flow and Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Nitao, J J

    2001-08-01

    This report was written as partial fulfillment of a subcontract from DOD/DOE Strategic Environmental Research and Development Program (SERDP) as part of a project directed by the U.S. Army Engineer Research and Development Center, Waterways Experiment Station (WES), Vicksburg, Mississippi. The report documents examples of field validation of the Non-isothermal Unsaturated-saturated Flow and Transport model (NUFT) code for environmental remediation, with emphasis on soil vapor extraction, and describes some of the modifications needed to integrate the code into the DOD Groundwater Modeling System (GMS, 2000). Note that this report highlights only a subset of the full capabilities of the NUFT code.

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

    Science.gov (United States)

    Phadte, D.; Patidar, C. B.; Pal, M. K.

    2017-04-01

    A relativistic Particle In Cell (PIC) code PARTDYN is developed for the beam dynamics simulation of z-continuous and bunched beams. The code is implemented in MATLAB using its MEX functionality which allows both ease of development as well higher performance similar to a compiled language like C. The beam dynamics calculations carried out by the code are compared with analytical results and with other well developed codes like PARMELA and BEAMPATH. The effect of finite number of simulation particles on the emittance growth of intense beams has been studied. Corrections to the RF cavity field expressions were incorporated in the code so that the fields could be calculated correctly. The deviations of the beam dynamics results between PARTDYN and BEAMPATH for a cavity driven in zero-mode have been discussed. The beam dynamics studies of the Low Energy Beam Transport (LEBT) using PARTDYN have been presented.

  4. NASA astronaut dosimetry: Implementation of scalable human phantoms and benchmark comparisons of deterministic versus Monte Carlo radiation transport

    Science.gov (United States)

    Bahadori, Amir Alexander

    Astronauts are exposed to a unique radiation environment in space. United States terrestrial radiation worker limits, derived from guidelines produced by scientific panels, do not apply to astronauts. Limits for astronauts have changed throughout the Space Age, eventually reaching the current National Aeronautics and Space Administration limit of 3% risk of exposure induced death, with an administrative stipulation that the risk be assured to the upper 95% confidence limit. Much effort has been spent on reducing the uncertainty associated with evaluating astronaut risk for radiogenic cancer mortality, while tools that affect the accuracy of the calculations have largely remained unchanged. In the present study, the impacts of using more realistic computational phantoms with size variability to represent astronauts with simplified deterministic radiation transport were evaluated. Next, the impacts of microgravity-induced body changes on space radiation dosimetry using the same transport method were investigated. Finally, dosimetry and risk calculations resulting from Monte Carlo radiation transport were compared with results obtained using simplified deterministic radiation transport. The results of the present study indicated that the use of phantoms that more accurately represent human anatomy can substantially improve space radiation dose estimates, most notably for exposures from solar particle events under light shielding conditions. Microgravity-induced changes were less important, but results showed that flexible phantoms could assist in optimizing astronaut body position for reducing exposures during solar particle events. Finally, little overall differences in risk calculations using simplified deterministic radiation transport and 3D Monte Carlo radiation transport were found; however, for the galactic cosmic ray ion spectra, compensating errors were observed for the constituent ions, thus exhibiting the need to perform evaluations on a particle

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

    Energy Technology Data Exchange (ETDEWEB)

    Mann, F.M.

    1998-01-26

    The Tank Waste Remediation System (TWRS) is responsible for the safe storage, retrieval, and disposal of waste currently being held in 177 underground tanks at the Hanford Site. In order to successfully carry out its mission, TWRS must perform environmental analyses describing the consequences of tank contents leaking from tanks and associated facilities during the storage, retrieval, or closure periods and immobilized low-activity tank waste contaminants leaving disposal facilities. Because of the large size of the facilities and the great depth of the dry zone (known as the vadose zone) underneath the facilities, sophisticated computer codes are needed to model the transport of the tank contents or contaminants. This document presents the code selection criteria for those vadose zone analyses (a subset of the above analyses) where the hydraulic properties of the vadose zone are constant in time the geochemical behavior of the contaminant-soil interaction can be described by simple models, and the geologic or engineered structures are complicated enough to require a two-or three dimensional model. Thus, simple analyses would not need to use the fairly sophisticated codes which would meet the selection criteria in this document. Similarly, those analyses which involve complex chemical modeling (such as those analyses involving large tank leaks or those analyses involving the modeling of contaminant release from glass waste forms) are excluded. The analyses covered here are those where the movement of contaminants can be relatively simply calculated from the moisture flow. These code selection criteria are based on the information from the low-level waste programs of the US Department of Energy (DOE) and of the US Nuclear Regulatory Commission as well as experience gained in the DOE Complex in applying these criteria. Appendix table A-1 provides a comparison between the criteria in these documents and those used here. This document does not define the models (that

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

    Science.gov (United States)

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

    2014-06-01

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

  7. ACCELERATING FUSION REACTOR NEUTRONICS MODELING BY AUTOMATIC COUPLING OF HYBRID MONTE CARLO/DETERMINISTIC TRANSPORT ON CAD GEOMETRY

    Energy Technology Data Exchange (ETDEWEB)

    Biondo, Elliott D [ORNL; Ibrahim, Ahmad M [ORNL; Mosher, Scott W [ORNL; Grove, Robert E [ORNL

    2015-01-01

    Detailed radiation transport calculations are necessary for many aspects of the design of fusion energy systems (FES) such as ensuring occupational safety, assessing the activation of system components for waste disposal, and maintaining cryogenic temperatures within superconducting magnets. Hybrid Monte Carlo (MC)/deterministic techniques are necessary for this analysis because FES are large, heavily shielded, and contain streaming paths that can only be resolved with MC. The tremendous complexity of FES necessitates the use of CAD geometry for design and analysis. Previous ITER analysis has required the translation of CAD geometry to MCNP5 form in order to use the AutomateD VAriaNce reducTion Generator (ADVANTG) for hybrid MC/deterministic transport. In this work, ADVANTG was modified to support CAD geometry, allowing hybrid (MC)/deterministic transport to be done automatically and eliminating the need for this translation step. This was done by adding a new ray tracing routine to ADVANTG for CAD geometries using the Direct Accelerated Geometry Monte Carlo (DAGMC) software library. This new capability is demonstrated with a prompt dose rate calculation for an ITER computational benchmark problem using both the Consistent Adjoint Driven Importance Sampling (CADIS) method an the Forward Weighted (FW)-CADIS method. The variance reduction parameters produced by ADVANTG are shown to be the same using CAD geometry and standard MCNP5 geometry. Significant speedups were observed for both neutrons (as high as a factor of 7.1) and photons (as high as a factor of 59.6).

  8. Monte Carlo Simulation of three dimensional Edwards Anderson model with multi-spin coding and parallel tempering using MPI and CUDA

    Science.gov (United States)

    Feng, Sheng; Fang, Ye; Tam, Ka-Ming; Thakur, Bhupender; Yun, Zhifeng; Tomko, Karen; Moreno, Juana; Ramanujam, Jagannathan; Jarrell, Mark

    2013-03-01

    The Edwards Anderson model is a typical example of random frustrated system. It has been a long standing problem in computational physics due to its long relaxation time. Some important properties of the low temperature spin glass phase are still poorly understood after decades of study. The recent advances of GPU computing provide a new opportunity to substantially improve the simulations. We developed an MPI-CUDA hybrid code with multi-spin coding for parallel tempering Monte Carlo simulation of Edwards Anderson model. Since the system size is relatively small, and a large number of parallel replicas and Monte Carlo moves are required, the problem suits well for modern GPUs with CUDA architecture. We use the code to perform an extensive simulation on the three-dimensional Edwards Anderson model with an external field. This work is funded by the NSF EPSCoR LA-SiGMA project under award number EPS-1003897. This work is partly done on the machines of Ohio Supercomputer Center.

  9. SQA of finite element method (FEM) codes used for analyses of pit storage/transport packages

    Energy Technology Data Exchange (ETDEWEB)

    Russel, E. [Lawrence Livermore National Lab., CA (United States)

    1997-11-01

    This report contains viewgraphs on the software quality assurance of finite element method codes used for analyses of pit storage and transport projects. This methodology utilizes the ISO 9000-3: Guideline for application of 9001 to the development, supply, and maintenance of software, for establishing well-defined software engineering processes to consistently maintain high quality management approaches.

  10. The neutron transport code DTF-Traca users manual and input data

    Energy Technology Data Exchange (ETDEWEB)

    Ahnert, C.

    1979-07-01

    This is a users manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data descriptions is given. The new options developed at JEN are included too. (Author) 18 refs.

  11. Comparison of Bonner sphere responses calculated by different Monte Carlo codes at energies between 1 MeV and 1 GeV – Potential impact on neutron dosimetry at energies higher than 20 MeV

    CERN Document Server

    Rühm, W; Pioch, C; Agosteo, S; Endo, A; Ferrarini, M; Rakhno, I; Rollet, S; Satoh, D; Vincke, H

    2014-01-01

    Bonner Spheres Spectrometry in its high-energy extended version is an established method to quantify neutrons at a wide energy range from several meV up to more than 1 GeV. In order to allow for quantitative measurements, the responses of the various spheres used in a Bonner Sphere Spectrometer (BSS) are usually simulated by Monte Carlo (MC) codes over the neutron energy range of interest. Because above 20 MeV experimental cross section data are scarce, intra-nuclear cascade (INC) and evaporation models are applied in these MC codes. It was suspected that this lack of data above 20 MeV may translate to differences in simulated BSS response functions depending on the MC code and nuclear models used, which in turn may add to the uncertainty involved in Bonner Sphere Spectrometry, in particular for neutron energies above 20 MeV. In order to investigate this issue in a systematic way, EURADOS (European Radiation Dosimetry Group) initiated an exercise where six groups having experience in neutron transport calcula...

  12. Adaptation of penelope Monte Carlo code system to the absorbed dose metrology: characterization of high energy photon beams and calculations of reference dosimeter correction factors; Adaptation du code Monte Carlo penelope pour la metrologie de la dose absorbee: caracterisation des faisceaux de photons X de haute energie et calcul de facteurs de correction de dosimetres de reference

    Energy Technology Data Exchange (ETDEWEB)

    Mazurier, J

    1999-05-28

    This thesis has been performed in the framework of national reference setting-up for absorbed dose in water and high energy photon beam provided with the SATURNE-43 medical accelerator of the BNM-LPRI (acronym for National Bureau of Metrology and Primary standard laboratory of ionising radiation). The aim of this work has been to develop and validate different user codes, based on PENELOPE Monte Carlo code system, to determine the photon beam characteristics and calculate the correction factors of reference dosimeters such as Fricke dosimeters and graphite calorimeter. In the first step, the developed user codes have permitted the influence study of different components constituting the irradiation head. Variance reduction techniques have been used to reduce the calculation time. The phase space has been calculated for 6, 12 and 25 MV at the output surface level of the accelerator head, then used for calculating energy spectra and dose distributions in the reference water phantom. Results obtained have been compared with experimental measurements. The second step has been devoted to develop an user code allowing calculation correction factors associated with both BNM-LPRI's graphite and Fricke dosimeters thanks to a correlated sampling method starting with energy spectra obtained in the first step. Then the calculated correction factors have been compared with experimental and calculated results obtained with the Monte Carlo EGS4 code system. The good agreement, between experimental and calculated results, leads to validate simulations performed with the PENELOPE code system. (author)

  13. Verification and Validation of Monte Carlo n-Particle Code 6 (MCNP6) with Neutron Protection Factor Measurements of an Iron Box

    Science.gov (United States)

    2014-03-27

    want to express my sincere love, respect, and admiration for my wife, who motivated and supported me throughout this long endeavor; this document ...widely utilized radiation transport code is MCNP. First created at Los Alamos National Laboratory ( LANL ) in 1957, the code simulated neutral...explanation of the current capabilities of MCNP will occur within the next chapter of this document ; however, it is important to note that MCNP

  14. Correlation of electron transport and photocatalysis of nanocrystalline clusters studied by Monte-Carlo continuity random walking.

    Science.gov (United States)

    Liu, Baoshun; Li, Ziqiang; Zhao, Xiujian

    2015-02-21

    In this research, Monte-Carlo Continuity Random Walking (MC-RW) model was used to study the relation between electron transport and photocatalysis of nano-crystalline (nc) clusters. The effects of defect energy disorder, spatial disorder of material structure, electron density, and interfacial transfer/recombination on the electron transport and the photocatalysis were studied. Photocatalytic activity is defined as 1/τ from a statistical viewpoint with τ being the electron average lifetime. Based on the MC-RW simulation, a clear physical and chemical "picture" was given for the photocatalytic kinetic analysis of nc-clusters. It is shown that the increase of defect energy disorder and material spatial structural disorder, such as the decrease of defect trap number, the increase of crystallinity, the increase of particle size, and the increase of inter-particle connection, can enhance photocatalytic activity through increasing electron transport ability. The increase of electron density increases the electron Fermi level, which decreases the activation energy for electron de-trapping from traps to extending states, and correspondingly increases electron transport ability and photocatalytic activity. Reducing recombination of electrons and holes can increase electron transport through the increase of electron density and then increases the photocatalytic activity. In addition to the electron transport, the increase of probability for electrons to undergo photocatalysis can increase photocatalytic activity through the increase of the electron interfacial transfer speed.

  15. Use of the ETA-1 reactor for the validation of the multi-group APOLLO2-MORET 5 code and the Monte Carlo continuous energy MORET 5 code

    Science.gov (United States)

    Leclaire, N.; Cochet, B.; Le Dauphin, F. X.; Haeck, W.; Jacquet, O.

    2014-06-01

    The present paper aims at providing experimental validation for the use of the MORET 5 code for advanced concepts of reactor involving thorium and heavy water. It therefore constitutes an opportunity to test and improve the thermal-scattering data of heavy water and also to test the recent implementation of probability tables in the MORET 5 code.

  16. CELLDOSE: A Monte Carlo code to assess electron dose distribution - S values for {sup 131}I in spheres of various sizes

    Energy Technology Data Exchange (ETDEWEB)

    Champion, C. [Univ Metz, Lab Phys Mol et Collis, Inst Phys, F-57078 Metz 3 (France); Zanotti-Fregonara, P. [Commissariat Energie Atom, DSV, I2BM, SHFJ, LIME, Orsay (France); Hindie, E [Hop St Louis, AP-HP, Paris (France); Hindie, E. [Imagerie Mol Diagnost et Ciblage Therapeut, Ecole Doctorale B2T, IUH, Paris, Univ Paris 07 (France)

    2008-07-01

    Monte Carlo simulation can be particularly suitable for modeling the microscopic distribution of energy received by normal tissues or cancer cells and for evaluating the relative merits of different radiopharmaceuticals. We used a new code, CELLDOSE, to assess electron dose for isolated spheres with radii varying from 2,500 {mu}m down to 0.05 {mu}m, in which {sup 131}I is homogeneously distributed. Methods: All electron emissions of {sup 131}I were considered,including the whole {beta}{sup -} {sup 131}I spectrum, 108 internal conversion electrons, and 21 Auger electrons. The Monte Carlo track-structure code used follows all electrons down to an energy threshold E-cutoff 7.4 eV. Results: Calculated S values were in good agreement with published analytic methods, lying in between reported results for all experimental points. Our S values were also close to other published data using a Monte Carlo code. Contrary to the latter published results, our results show that dose distribution inside spheres is not homogeneous, with the dose at the outmost layer being approximately half that at the center. The fraction of electron energy retained within the spheres decreased with decreasing radius (r): 87.1 % for r 2,500 {mu}m, 8.73% for r 50 {mu}m, and 1.18% for r 5 {mu}m. Thus, a radioiodine concentration that delivers a dose of 100 Gy to a micro-metastasis of 2,500 {mu}m radius would deliver 10 Gy in a cluster of 50 {mu}m and only 1.4 Gy in an isolated cell. The specific contribution from Auger electrons varied from 0.25% for the largest sphere up to 76.8% for the smallest sphere. Conclusion: The dose to a tumor cell will depend on its position in a metastasis. For the treatment of very small metastases, {sup 131}I may not be the isotope of choice. When trying to kill isolated cells or a small cluster of cells with {sup 131}I, it is important to get the iodine as close as possible to the nucleus to get the enhancement factor from Auger electrons. The Monte Carlo code

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

    Science.gov (United States)

    Hammond, G. E.; Andre, B.; Bisht, G.; Johnson, T.; Karra, S.; Lichtner, P. C.; Mills, R. T.

    2013-12-01

    Open-source software development has become increasingly popular in recent years. Open-source encourages collaborative and transparent software development and promotes unlimited free redistribution of source code to the public. Open-source development is good for science as it reveals implementation details that are critical to scientific reproducibility, but generally excluded from journal publications. In addition, research funds that would have been spent on licensing fees can be redirected to code development that benefits more scientists. In 2006, the developers of PFLOTRAN open-sourced their code under the U.S. Department of Energy SciDAC-II program. Since that time, the code has gained popularity among code developers and users from around the world seeking to employ PFLOTRAN to simulate thermal, hydraulic, mechanical and biogeochemical processes in the Earth's surface/subsurface environment. PFLOTRAN is a massively-parallel subsurface reactive multiphase flow and transport simulator designed from the ground up to run efficiently on computing platforms ranging from the laptop to leadership-class supercomputers, all from a single code base. The code employs domain decomposition for parallelism and is founded upon the well-established and open-source parallel PETSc and HDF5 frameworks. PFLOTRAN leverages modern Fortran (i.e. Fortran 2003-2008) in its extensible object-oriented design. The use of this progressive, yet domain-friendly programming language has greatly facilitated collaboration in the code's software development. Over the past year, PFLOTRAN's top-level data structures were refactored as Fortran classes (i.e. extendible derived types) to improve the flexibility of the code, ease the addition of new process models, and enable coupling to external simulators. For instance, PFLOTRAN has been coupled to the parallel electrical resistivity tomography code E4D to enable hydrogeophysical inversion while the same code base can be used as a third

  18. A fast GPU-based Monte Carlo simulation of proton transport with detailed modeling of non-elastic interactions

    CERN Document Server

    Tseung, H Wan Chan; Beltran, C

    2014-01-01

    Purpose: Very fast Monte Carlo (MC) simulations of proton transport have been implemented recently on GPUs. However, these usually use simplified models for non-elastic (NE) proton-nucleus interactions. Our primary goal is to build a GPU-based proton transport MC with detailed modeling of elastic and NE collisions. Methods: Using CUDA, we implemented GPU kernels for these tasks: (1) Simulation of spots from our scanning nozzle configurations, (2) Proton propagation through CT geometry, considering nuclear elastic scattering, multiple scattering, and energy loss straggling, (3) Modeling of the intranuclear cascade stage of NE interactions, (4) Nuclear evaporation simulation, and (5) Statistical error estimates on the dose. To validate our MC, we performed: (1) Secondary particle yield calculations in NE collisions, (2) Dose calculations in homogeneous phantoms, (3) Re-calculations of head and neck plans from a commercial treatment planning system (TPS), and compared with Geant4.9.6p2/TOPAS. Results: Yields, en...

  19. An improved Monte Carlo study of coherent scattering effects of low energy charged particle transport in Percus-Yevick liquids

    CERN Document Server

    Tattersall, W J; Boyle, G J; White, R D

    2015-01-01

    We generalize a simple Monte Carlo (MC) model for dilute gases to consider the transport behavior of positrons and electrons in Percus-Yevick model liquids under highly non-equilibrium conditions, accounting rigorously for coherent scattering processes. The procedure extends an existing technique [Wojcik and Tachiya, Chem. Phys. Lett. 363, 3--4 (1992)], using the static structure factor to account for the altered anisotropy of coherent scattering in structured material. We identify the effects of the approximation used in the original method, and develop a modified method that does not require that approximation. We also present an enhanced MC technique that has been designed to improve the accuracy and flexibility of simulations in spatially-varying electric fields. All of the results are found to be in excellent agreement with an independent multi-term Boltzmann equation solution, providing benchmarks for future transport models in liquids and structured systems.

  20. 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.

  1. 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.

  2. A multi-agent quantum Monte Carlo model for charge transport: Application to organic field-effect transistors

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, Thilo; Jäger, Christof M. [Department of Chemistry and Pharmacy, Computer-Chemistry-Center and Interdisciplinary Center for Molecular Materials, Friedrich-Alexander-Universität Erlangen-Nürnberg, Nägelsbachstrasse 25, 91052 Erlangen (Germany); Jordan, Meredith J. T. [School of Chemistry, University of Sydney, Sydney, NSW 2006 (Australia); Clark, Timothy, E-mail: tim.clark@fau.de [Department of Chemistry and Pharmacy, Computer-Chemistry-Center and Interdisciplinary Center for Molecular Materials, Friedrich-Alexander-Universität Erlangen-Nürnberg, Nägelsbachstrasse 25, 91052 Erlangen (Germany); Centre for Molecular Design, University of Portsmouth, Portsmouth PO1 2DY (United Kingdom)

    2015-07-28

    We have developed a multi-agent quantum Monte Carlo model to describe the spatial dynamics of multiple majority charge carriers during conduction of electric current in the channel of organic field-effect transistors. The charge carriers are treated by a neglect of diatomic differential overlap Hamiltonian using a lattice of hydrogen-like basis functions. The local ionization energy and local electron affinity defined previously map the bulk structure of the transistor channel to external potentials for the simulations of electron- and hole-conduction, respectively. The model is designed without a specific charge-transport mechanism like hopping- or band-transport in mind and does not arbitrarily localize charge. An electrode model allows dynamic injection and depletion of charge carriers according to source-drain voltage. The field-effect is modeled by using the source-gate voltage in a Metropolis-like acceptance criterion. Although the current cannot be calculated because the simulations have no time axis, using the number of Monte Carlo moves as pseudo-time gives results that resemble experimental I/V curves.

  3. Monte Carlo simulation and Boltzmann equation analysis of non-conservative positron transport in H{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Bankovic, A., E-mail: ana.bankovic@gmail.com [Institute of Physics, University of Belgrade, Pregrevica 118, 11080 Belgrade (Serbia); Dujko, S. [Institute of Physics, University of Belgrade, Pregrevica 118, 11080 Belgrade (Serbia); Centrum Wiskunde and Informatica (CWI), P.O. Box 94079, 1090 GB Amsterdam (Netherlands); ARC Centre for Antimatter-Matter Studies, School of Engineering and Physical Sciences, James Cook University, Townsville, QLD 4810 (Australia); White, R.D. [ARC Centre for Antimatter-Matter Studies, School of Engineering and Physical Sciences, James Cook University, Townsville, QLD 4810 (Australia); Buckman, S.J. [ARC Centre for Antimatter-Matter Studies, Australian National University, Canberra, ACT 0200 (Australia); Petrovic, Z.Lj. [Institute of Physics, University of Belgrade, Pregrevica 118, 11080 Belgrade (Serbia)

    2012-05-15

    This work reports on a new series of calculations of positron transport properties in molecular hydrogen under the influence of spatially homogeneous electric field. Calculations are performed using a Monte Carlo simulation technique and multi term theory for solving the Boltzmann equation. Values and general trends of the mean energy, drift velocity and diffusion coefficients as a function of the reduced electric field E/n{sub 0} are reported here. Emphasis is placed on the explicit and implicit effects of positronium (Ps) formation on the drift velocity and diffusion coefficients. Two important phenomena arise; first, for certain regions of E/n{sub 0} the bulk and flux components of the drift velocity and longitudinal diffusion coefficient are markedly different, both qualitatively and quantitatively. Second, and contrary to previous experience in electron swarm physics, there is negative differential conductivity (NDC) effect in the bulk drift velocity component with no indication of any NDC for the flux component. In order to understand this atypical manifestation of the drift and diffusion of positrons in H{sub 2} under the influence of electric field, the spatially dependent positron transport properties such as number of positrons, average energy and velocity and spatially resolved rate for Ps formation are calculated using a Monte Carlo simulation technique. The spatial variation of the positron average energy and extreme skewing of the spatial profile of positron swarm are shown to play a central role in understanding the phenomena.

  4. Optimizing light transport in scintillation crystals for time-of-flight PET: an experimental and optical Monte Carlo simulation study.

    Science.gov (United States)

    Berg, Eric; Roncali, Emilie; Cherry, Simon R

    2015-06-01

    Achieving excellent timing resolution in gamma ray detectors is crucial in several applications such as medical imaging with time-of-flight positron emission tomography (TOF-PET). Although many factors impact the overall system timing resolution, the statistical nature of scintillation light, including photon production and transport in the crystal to the photodetector, is typically the limiting factor for modern scintillation detectors. In this study, we investigated the impact of surface treatment, in particular, roughening select areas of otherwise polished crystals, on light transport and timing resolution. A custom Monte Carlo photon tracking tool was used to gain insight into changes in light collection and timing resolution that were observed experimentally: select roughening configurations increased the light collection up to 25% and improved timing resolution by 15% compared to crystals with all polished surfaces. Simulations showed that partial surface roughening caused a greater number of photons to be reflected towards the photodetector and increased the initial rate of photoelectron production. This study provides a simple method to improve timing resolution and light collection in scintillator-based gamma ray detectors, a topic of high importance in the field of TOF-PET. Additionally, we demonstrated utility of our Monte Carlo simulation tool to accurately predict the effect of altering crystal surfaces on light collection and timing resolution.

  5. Assessment of shielding analysis methods, codes, and data for spent fuel transport/storage applications. [Radiation dose rates from shielded spent fuels and high-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Parks, C.V.; Broadhead, B.L.; Hermann, O.W.; Tang, J.S.; Cramer, S.N.; Gauthey, J.C.; Kirk, B.L.; Roussin, R.W.

    1988-07-01

    This report provides a preliminary assessment of the computational tools and existing methods used to obtain radiation dose rates from shielded spent nuclear fuel and high-level radioactive waste (HLW). Particular emphasis is placed on analysis tools and techniques applicable to facilities/equipment designed for the transport or storage of spent nuclear fuel or HLW. Applications to cask transport, storage, and facility handling are considered. The report reviews the analytic techniques for generating appropriate radiation sources, evaluating the radiation transport through the shield, and calculating the dose at a desired point or surface exterior to the shield. Discrete ordinates, Monte Carlo, and point kernel methods for evaluating radiation transport are reviewed, along with existing codes and data that utilize these methods. A literature survey was employed to select a cadre of codes and data libraries to be reviewed. The selection process was based on specific criteria presented in the report. Separate summaries were written for several codes (or family of codes) that provided information on the method of solution, limitations and advantages, availability, data access, ease of use, and known accuracy. For each data library, the summary covers the source of the data, applicability of these data, and known verification efforts. Finally, the report discusses the overall status of spent fuel shielding analysis techniques and attempts to illustrate areas where inaccuracy and/or uncertainty exist. The report notes the advantages and limitations of several analysis procedures and illustrates the importance of using adequate cross-section data sets. Additional work is recommended to enable final selection/validation of analysis tools that will best meet the US Department of Energy's requirements for use in developing a viable HLW management system. 188 refs., 16 figs., 27 tabs.

  6. Monte Carlo transport simulation for a long counter neutron detector employed as a cosmic rays induced neutron monitor at ground level

    Energy Technology Data Exchange (ETDEWEB)

    Pazianotto, Mauricio Tizziani; Carlson, Brett Vern [Instituto Tecnologico de Aeronautica (ITA), Sao Jose dos Campos, SP (Brazil); Federico, Claudio Antonio; Goncalez, Odair Lelis [Centro Tecnico Aeroespacial (CTA), Sao Jose dos Campos, SP (Brazil). Instituto de Estudos Avancados

    2011-07-01

    Full text: Great effort is required to understand better the cosmic radiation (CR) dose received by sensitive equipment, on-board computers and aircraft crew members at Brazil airspace, because there is a large area of South America and Brazil subject to the South Atlantic Anomaly (SAA). High energy neutrons are produced by interactions between primary cosmic ray and atmospheric atoms, and also undergo moderation resulting in a wider spectrum of energy ranging from thermal energies (0:025eV ) to energies of several hundreds of MeV. Measurements of the cosmic radiation dose on-board aircrafts need to be followed with an integral flow monitor on the ground level in order to register CR intensity variations during the measurements. The Long Counter (LC) neutron detector was designed as a directional neutron flux meter standard because it presents fairly constant response for energy under 10MeV. However we would like to use it as a ground based neutron monitor for cosmic ray induced neutron spectrum (CRINS) that presents an isotropic fluency and a wider spectrum of energy. The LC was modeled and tested using a Monte Carlo transport simulation for irradiations with known neutron sources ({sup 241}Am-Be and {sup 251}Cf) as a benchmark. Using this geometric model its efficiency was calculated to CRINS isotropic flux, introducing high energy neutron interactions models. The objective of this work is to present the model for simulation of the isotropic neutron source employing the MCNPX code (Monte Carlo N-Particle eXtended) and then access the LC efficiency to compare it with experimental results for cosmic ray neutrons measures on ground level. (author)

  7. Radial transport dynamics studies of SMBI with a newly developed TPSMBI code

    Science.gov (United States)

    Wang, Ya-Hui; Guo, Wen-Feng; Wang, Zhan-Hui; Ren, Qi-Long; Sun, Ai-Ping; Xu, Min; Wang, Ai-Ke; Xiang, Nong

    2016-10-01

    In tokamak plasma fueling, supersonic molecule beam injection (SMBI) with a higher fueling efficiency and a deeper penetration depth than the traditional gas puffing method has been developed and widely applied to many tokamak devices. It is crucial to study the transport dynamics of SMBI to improve its fueling efficiency, especially in the high confinement regime. A new one-dimensional (1D) code of TPSMBI has also been developed recently based on a six-field SMBI model in cylindrical coordinate. It couples plasma density and heat radial transport equations together with neutral density transport equations for both molecules and atoms and momentum radial transport equations for molecules. The dominant particle collisional interactions between plasmas and neutrals, such as molecule dissociation, atom ionization and charge-exchange effects, are included in the model. The code is verified to be correct with analytical solutions and also benchmarked well with the trans-neut module of BOUT++ code. Time-dependent radial transport dynamics and mean profile evolution are studied during SMBI with the TPSMBI code in both slab and cylindrical coordinates. Along the SMBI path, plasma density increases due to particle fuelling, while plasma temperature decreases due to heat cooling. Being different from slab coordinate, the curvature effect leads to larger front densities of molecule and atom during SMBI in cylindrical coordinate simulation. Project supported by the National Natural Science Foundation of China (Grant Nos. 11575055, 11375053, and 11475219) and the National Magnetic Confinement Fusion Science Program of China (Grant Nos. 2013GB111005, 2014GB108004, and 2015GB110001).

  8. New Dosimetric Interpretation of the DV50 Vessel-Steel Experiment Irradiated in the OSIRIS MTR Reactor Using the Monte-Carlo Code TRIPOLI-4®

    Directory of Open Access Journals (Sweden)

    Malouch Fadhel

    2016-01-01

    Full Text Available An irradiation program DV50 was carried out from 2002 to 2006 in the OSIRIS material testing reactor (CEA-Saclay center to assess the pressure vessel steel toughness curve for a fast neutron fluence (E > 1 MeV equivalent to a French 900-MWe PWR lifetime of 50 years. This program allowed the irradiation of 120 specimens out of vessel steel, subdivided in two successive irradiations DV50 n∘1 and DV50 n∘2. To measure the fast neutron fluence (E > 1 MeV received by specimens after each irradiation, sample holders were equipped with activation foils that were withdrawn at the end of irradiation for activity counting and processing. The fast effective cross-sections used in the dosimeter processing were determined with a specific calculation scheme based on the Monte-Carlo code TRIPOLI-3 (and the nuclear data ENDF/B-VI and IRDF-90. In order to put vessel-steel experiments at the same standard, a new dosimetric interpretation of the DV50 experiment has been performed by using the Monte-Carlo code TRIPOLI-4 and more recent nuclear data (JEFF3.1.1 and IRDF-2002. This paper presents a comparison of previous and recent calculations performed for the DV50 vessel-steel experiment to assess the impact on the dosimetric interpretation.

  9. Emission from Very Small Grains and PAH Molecules in Monte Carlo Radiation Transfer Codes: Application to the Edge-On Disk of Gomez's Hamburger

    Science.gov (United States)

    Wood, Kenneth; Whitney, Barbara A.; Robitaille, Thomas; Draine, Bruce T.

    2008-12-01

    We have modeled optical to far-infrared images, photometry, and spectroscopy of the object known as Gomez's Hamburger. We reproduce the images and spectrum with an edge-on disk of mass 0.3 M⊙ and radius 1600 AU, surrounding an A0 III star at a distance of 280 pc. Our mass estimate is in excellent agreement with recent CO observations. However, our distance determination is more than an order of magnitude smaller than previous analyses, which inaccurately interpreted the optical spectrum. To accurately model the infrared spectrum we have extended our Monte Carlo radiation transfer codes to include emission from polycyclic aromatic hydrocarbon (PAH) molecules and very small grains (VSG). We do this using precomputed PAH/VSG emissivity files for a wide range of values of the mean intensity of the exciting radiation field. When Monte Carlo energy packets are absorbed by PAHs/VSGs, we reprocess them to other wavelengths by sampling from the emissivity files, thus simulating the absorption and reemission process without reproducing lengthy computations of statistical equilibrium, excitation, and de-excitation in the complex many-level molecules. Using emissivity lookup tables in our Monte Carlo codes gives us the flexibility to use the latest grain physics calculations of PAH/VSG emissivity and opacity that are being continually updated in the light of higher resolution infrared spectra. We find our approach gives a good representation of the observed PAH spectrum from the disk of Gomez's Hamburger. Our models also indicate that the PAHs/VSGs in the disk have a larger scale height than larger radiative equilibrium grains, providing evidence for dust coagulation and settling to the midplane.

  10. Multilevel Monte Carlo methods using ensemble level mixed MsFEM for two-phase flow and transport simulations

    KAUST Repository

    Efendiev, Yalchin R.

    2013-08-21

    In this paper, we propose multilevel Monte Carlo (MLMC) methods that use ensemble level mixed multiscale methods in the simulations of multiphase flow and transport. The contribution of this paper is twofold: (1) a design of ensemble level mixed multiscale finite element methods and (2) a novel use of mixed multiscale finite element methods within multilevel Monte Carlo techniques to speed up the computations. The main idea of ensemble level multiscale methods is to construct local multiscale basis functions that can be used for any member of the ensemble. In this paper, we consider two ensemble level mixed multiscale finite element methods: (1) the no-local-solve-online ensemble level method (NLSO); and (2) the local-solve-online ensemble level method (LSO). The first approach was proposed in Aarnes and Efendiev (SIAM J. Sci. Comput. 30(5):2319-2339, 2008) while the second approach is new. Both mixed multiscale methods use a number of snapshots of the permeability media in generating multiscale basis functions. As a result, in the off-line stage, we construct multiple basis functions for each coarse region where basis functions correspond to different realizations. In the no-local-solve-online ensemble level method, one uses the whole set of precomputed basis functions to approximate the solution for an arbitrary realization. In the local-solve-online ensemble level method, one uses the precomputed functions to construct a multiscale basis for a particular realization. With this basis, the solution corresponding to this particular realization is approximated in LSO mixed multiscale finite element method (MsFEM). In both approaches, the accuracy of the method is related to the number of snapshots computed based on different realizations that one uses to precompute a multiscale basis. In this paper, ensemble level multiscale methods are used in multilevel Monte Carlo methods (Giles 2008a, Oper.Res. 56(3):607-617, b). In multilevel Monte Carlo methods, more accurate

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

    Energy Technology Data Exchange (ETDEWEB)

    Saegusa, T. [Central Research Inst. of Electric Power Industry, Abiko (Japan); Arai, T. [Central Research Inst. of Electric Power Industry, Yokosuka (Japan); Hirose, M. [Nuclear Fuel Transport Co., Ltd., Tokyo (Japan); Kobayashi, T. [Nippon Chuzo, Kawasaki (Japan); Tezuka, Y. [Mitsubishi Materials Co., Tokyo (Japan); Urabe, N. [Kokan Keisoku K. K., Kawasaki (Japan); Hueggenberg, R. [GNB, Essen (Germany)

    2004-07-01

    The current Rules for Construction of ''Containment Systems for Storage and Transport Packagings of Spent Nuclear Fuel and High Level Radioactive Material and Waste'' of Division 3 in Section III of ASME Code (2001 Edition) does not include ductile cast iron in its list of materials permitted for use. The Rules specify required fracture toughness values of ferritic steel material for nominal wall thickness 5/8 to 12 inches (16 to 305 mm). New rule for ductile cast iron for transport packaging of which wall thickness is greater than 12 inches (305mm) is required.

  12. Development of a tritium transport analysis code for the LMFBR system

    Energy Technology Data Exchange (ETDEWEB)

    Iizawa, Katsuyuki; Torii, Tatsuo [Japan Nuclear Cycle Development Inst., Tsuruga Head Office, Tsuruga, Fukui (Japan)

    2001-03-01

    A tritium transport analysis code for the LMFBR system, TTT code, has been developed and validated using data from a power rising test conducted at Monju in 1995. The behavior of tritium during future long-term full power operation of Monju has been estimated. The TTT code was created from the tritium and hydrogen transport model devised by R. Kumar and ANL. Actual data from some plants has been used to improve the code. In this study, we used data from Monju to increase the accuracy of the calculated to measured ratio, the C/E ratio. As a result of the study, we were able to: 1. show that the calculated tritium concentration distribution and the change in the primary and secondary sodium, steam and water correlated sufficiently closely with the measured, C/E ratio of 1.1; 2. propose a transport model between sodium and the cover gas system taking into account the mechanisms affecting the partial pressure difference and the isotopic exchange of H and H3; 3. examine the considerable effect of the hydrogen source within the sodium cooling system of Monju on tritium behavior and clarify the characteristics at the initial stage of plant; 4. estimate the tritium transport and distribution for the long-term full power operation of Monju. The tritium release from the core will be 7,400 TBq during 30 years of operation. The primary and secondary cold trap will capture 99% of this and 1% or less will be released to the environment as gaseous radioactive waste from stack and its drainage water from SG; and 5. compare the best fitted tritium source rates from cores in Phenix and Monju and estimate the major release from Monju's helium bond closed type control rods. (author)

  13. Monte Carlo transport model comparison with 1A GeV accelerated iron experiment: heavy-ion shielding evaluation of NASA space flight-crew foodstuff

    Science.gov (United States)

    Stephens, D. L.; Townsend, L. W.; Miller, J.; Zeitlin, C.; Heilbronn, L.

    Deep-space manned flight as a reality depends on a viable solution to the radiation problem. Both acute and chronic radiation health threats are known to exist, with solar particle events as an example of the former and galactic cosmic rays (GCR) of the latter. In this experiment Iron ions of 1A GeV are used to simulate GCR and to determine the secondary radiation field created as the GCR-like particles interact with a thick target. A NASA prepared food pantry locker was subjected to the iron beam and the secondary fluence recorded. A modified version of the Monte Carlo heavy ion transport code developed by Zeitlin at LBNL is compared with experimental fluence. The foodstuff is modeled as mixed nuts as defined by the 71 st edition of the Chemical Rubber Company (CRC) Handbook of Physics and Chemistry. The results indicate a good agreement between the experimental data and the model. The agreement between model and experiment is determined using a linear fit to ordered pairs of data. The intercept is forced to zero. The slope fit is 0.825 and the R 2 value is 0.429 over the resolved fluence region. The removal of an outlier, Z=14, gives values of 0.888 and 0.705 for slope and R 2 respectively.

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

    Energy Technology Data Exchange (ETDEWEB)

    Coste-Delclaux, M

    2006-03-15

    This document describes the improvements carried out for modelling the self-shielding phenomenon in the multigroup transport code APOLLO2. They concern the space and energy treatment of the slowing-down equation, the setting up of quadrature formulas to calculate reaction rates, the setting-up of a method that treats directly a resonant mixture and the development of a sub-group method. We validate these improvements either in an elementary or in a global way. Now, we obtain, more accurate multigroup reaction rates and we are able to carry out a reference self-shielding calculation on a very fine multigroup mesh. To end, we draw a conclusion and give some prospects on the remaining work. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Ganapol, B.D.; Kornreich, D.E. [Univ. of Arizona, Tucson, AZ (United States). Dept. of Nuclear Engineering

    1997-07-01

    Because of the requirement of accountability and quality control in the scientific world, a demand for high-quality analytical benchmark calculations has arisen in the neutron transport community. The intent of these benchmarks is to provide a numerical standard to which production neutron transport codes may be compared in order to verify proper operation. The overall investigation as modified in the second year renewal application includes the following three primary tasks. Task 1 on two dimensional neutron transport is divided into (a) single medium searchlight problem (SLP) and (b) two-adjacent half-space SLP. Task 2 on three-dimensional neutron transport covers (a) point source in arbitrary geometry, (b) single medium SLP, and (c) two-adjacent half-space SLP. Task 3 on code verification, includes deterministic and probabilistic codes. The primary aim of the proposed investigation was to provide a suite of comprehensive two- and three-dimensional analytical benchmarks for neutron transport theory applications. This objective has been achieved. The suite of benchmarks in infinite media and the three-dimensional SLP are a relatively comprehensive set of one-group benchmarks for isotropically scattering media. Because of time and resource limitations, the extensions of the benchmarks to include multi-group and anisotropic scattering are not included here. Presently, however, enormous advances in the solution for the planar Green`s function in an anisotropically scattering medium have been made and will eventually be implemented in the two- and three-dimensional solutions considered under this grant. Of particular note in this work are the numerical results for the three-dimensional SLP, which have never before been presented. The results presented were made possible only because of the tremendous advances in computing power that have occurred during the past decade.

  16. ARCHER, a New Monte Carlo Software Tool for Emerging Heterogeneous Computing Environments

    Science.gov (United States)

    Xu, X. George; Liu, Tianyu; Su, Lin; Du, Xining; Riblett, Matthew; Ji, Wei; Gu, Deyang; Carothers, Christopher D.; Shephard, Mark S.; Brown, Forrest B.; Kalra, Mannudeep K.; Liu, Bob

    2014-06-01

    The Monte Carlo radiation transport community faces a number of challenges associated with peta- and exa-scale computing systems that rely increasingly on heterogeneous architectures involving hardware accelerators such as GPUs. Existing Monte Carlo codes and methods must be strategically upgraded to meet emerging hardware and software needs. In this paper, we describe the development of a software, called ARCHER (Accelerated Radiation-transport Computations in Heterogeneous EnviRonments), which is designed as a versatile testbed for future Monte Carlo codes. Preliminary results from five projects in nuclear engineering and medical physics are presented.

  17. Comparison of Fuel Temperature Coefficients of PWR UO{sub 2} Fuel from Monte Carlo Codes (MCNP6.1 and KENO6)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung-O; Roh, Gyuhong; Lee, Byungchul [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    As a result, there was a difference within about 300-400 pcm between keff values at each enrichment due to the difference of codes and nuclear data used in the evaluations. The FTC was changed to be less negative with the increase of uranium enrichment, and it followed the form of asymptotic curve. However, it is necessary to perform additional study for investigating what factor causes the differences more than two standard deviation (2σ) among the FTCs at partial enrichment region. The interaction probability of incident neutron with nuclear fuel is depended on the relative velocity between the neutron and the target nuclei. The Fuel Temperature Coefficient (FTC) is defined as the change of Doppler effect with respect to the change in fuel temperature without any other change such as moderator temperature, moderator density, etc. In this study, the FTCs for UO{sub 2} fuel were evaluated by using MCNP6.1 and KENO6 codes based on a Monte Carlo method. In addition, the latest neutron cross-sections (ENDF/B-VI and VII) were applied to analyze the effect of these data on the evaluation of FTC, and nuclear data used in MCNP calculations were generated from the makxsf code. An evaluation of the Doppler effect and FTC for UO{sub 2} fuel widely used in PWR was conducted using MCNP6.1 and KENO6 codes. The ENDF/B-VI and VII were also applied to analyze what effect these data has on those evaluations. All cross-sections needed for MCNP calculation were produced using makxsf code. The calculation models used in the evaluations were based on the typical PWR UO{sub 2} lattice.

  18. Treatment of patient-dependent beam modifiers in photon treatments by the Monte Carlo dose calculation code PEREGRINE

    Energy Technology Data Exchange (ETDEWEB)

    Schach von Wittenau, A.E.; Cox, L.J.; Bergstrom, P.M. Jr.; Hornstein, S.M. [Lawrence Livermore National Lab., CA (United States); Mohan, R.; Libby, B.; Wu, Q. [Medical Coll. of Virginia, Richmond, VA (United States); Lovelock, D.M.J. [Memorial Sloan-Kettering Cancer Center, New York, NY (United States)

    1997-03-01

    The goal of the PEREGRINE Monte Carlo Dose Calculation Project is to deliver a Monte Carlo package that is both accurate and sufficiently fast for routine clinical use. One of the operational requirements for photon-treatment plans is a fast, accurate method of describing the photon phase-space distribution at the surface of the patient. The open-field case is computationally the most tractable; we know, a priori, for a given machine and energy, the locations and compositions of the relevant accelerator components (i.e., target, primary collimator, flattening filter, and monitor chamber). Therefore, we can precalculate and store the expected photon distributions. For any open-field treatment plan, we then evaluate these existing photon phase-space distributions at the patient`s surface, and pass the obtained photons to the dose calculation routines within PEREGRINE. We neglect any effect of the intervening air column, including attenuation of the photons and production of contaminant electrons. In principle, for treatment plans requiring jaws, blocks, and wedges, we could precalculate and store photon phase-space distributions for various combinations of field sizes and wedges. This has the disadvantage that we would have to anticipate those combinations and that subsequently PEREGRINE would not be able to treat other plans. Therefore, PEREGRINE tracks photons through the patient-dependent beam modifiers. The geometric and physics methods used to do this are described here. 4 refs., 8 figs.

  19. Methodology to resolve the transport equation with the discrete ordinates code TORT into the IPEN/MB-01 reactor

    OpenAIRE

    Bernal García, Álvaro; Abarca Giménez, Agustín; Barrachina Celda, Teresa María; Miró Herrero, Rafael

    2014-01-01

    This is an Accepted Manuscript of an article published by Taylor & Francis in International Journal of Computer Mathematics in 2014, available online: http://www.tandfonline.com/10.1080/00207160.2013.799668 Resolution of the steady-state Neutron Transport Equation in a nuclear pool reactor is usually achieved by means of two different numerical methods: Monte Carlo (stochastic) and Discrete Ordinates (deterministic). The Discrete Ordinates method solves the Neutron Transport Equation for a...

  20. An improved empirical approach to introduce quantization effects in the transport direction in multi-subband Monte Carlo simulations

    Science.gov (United States)

    Palestri, P.; Lucci, L.; Dei Tos, S.; Esseni, D.; Selmi, L.

    2010-05-01

    In this paper we propose and validate a simple approach to empirically account for quantum effects in the transport direction of MOS transistors (i.e. source and drain tunneling and delocalized nature of the carrier wavepacket) in multi-subband Monte Carlo simulators, that already account for quantization in the direction normal to the semiconductor-oxide interface by solving the 1D Schrödinger equation in each section of the device. The model has been validated and calibrated against ballistic non-equilibrium Green's function simulations over a wide range of gate lengths, voltage biases and temperatures. The proposed model has just one adjustable parameter and our results show that it can achieve a good agreement with the NEGF approach.

  1. Monte Carlo simulation on electron transport in Si sub 1 sub - sub y C sub y alloy layers

    CERN Document Server

    Ihm, S H; Lee, C H; Lee, H J; Kim, J Y; Chun, S K

    1999-01-01

    We investigated electron transport in strained Si sub 1 sub - sub y C sub y alloy layers grown on Si(100) substrates using the Monte Carlo simulation. The electron mobility higher than that of bulk Si over a wide range of temperatures from 40 K to 300 K is mainly attributed to the valley splitting induced by the tensile strain in the Si sub 1 sub - sub y C sub y layer. For lower temperatures less than 100 K the mobility increases sharply depending on the carbon fraction up to about 0.6%. Beyond the fraction, however, it keeps almost constant regardless of increasing the carbon fraction. On the other hand, we observe a monotonic mobility increase with increasing the carbon for a higher temperature regime.

  2. Monte Carlo Simulation of Light Transport in Five-Layered Skin Tissue

    Institute of Scientific and Technical Information of China (English)

    XUE Ling-Ling; ZHANG Chun-Ping; WANG Xin-Yu; ZHU Ming-Yao; ZHANG Lian-Shun; CHI Rong-Hua; ZHANG Jian-Dong; ZHANG Guang-Yin

    2000-01-01

    The light propagation and distribution in skin tissue is studied by using Monte Carlo technique. The radially resolved diffuse reflectance R and transmittance T vs radius r, angularly resolved R and Tvs the exiting angle of the photon, absorption energy density A and internal fiuence F vs r and z are simulated. Our results reveal that the light distribution for Gaussian beam is more centralized and its change is more rapid than those of circularly flat beam under the same incident energy and radius, no matter what R and T or A and F are. In addition,except that R(r) for circularly flat beam needs to be fitted by 15-order curve, the others can be fitted by 5-order or 6-order curve.

  3. Contaminant transport in fracture networks with heterogeneous rock matrices. The Picnic code

    Energy Technology Data Exchange (ETDEWEB)

    Barten, Werner [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland); Robinson, Peter C. [QuantiSci Limited, Henley-on-Thames (United Kingdom)

    2001-02-01

    In the context of safety assessment of radioactive waste repositories, complex radionuclide transport models covering key safety-relevant processes play a major role. In recent Swiss safety assessments, such as Kristallin-I, an important drawback was the limitation in geosphere modelling capability to account for geosphere heterogeneities. In marked contrast to this limitation in modelling capabilities, great effort has been put into investigating the heterogeneity of the geosphere as it impacts on hydrology. Structural geological methods have been used to look at the geometry of the flow paths on a small scale and the diffusion and sorption properties of different rock materials have been investigated. This huge amount of information could however be only partially applied in geosphere transport modelling. To make use of these investigations the 'PICNIC project' was established as a joint cooperation of PSI/Nagra and QuantiSci to provide a new geosphere transport model for Swiss safety assessment of radioactive waste repositories. The new transport code, PICNIC, can treat all processes considered in the older geosphere model RANCH MD generally used in the Kristallin-I study and, in addition, explicitly accounts for the heterogeneity of the geosphere on different spatial scales. The effects and transport phenomena that can be accounted for by PICNIC are a combination of (advective) macro-dispersion due to transport in a network of conduits (legs), micro-dispersion in single legs, one-dimensional or two-dimensional matrix diffusion into a wide range of homogeneous and heterogeneous rock matrix geometries, linear sorption of nuclides in the flow path and the rock matrix and radioactive decay and ingrowth in the case of nuclide chains. Analytical and numerical Laplace transformation methods are integrated in a newly developed hierarchical linear response concept to efficiently account for the transport mechanisms considered which typically act on extremely

  4. The effect of load imbalances on the performance of Monte Carlo algorithms in LWR analysis

    Energy Technology Data Exchange (ETDEWEB)

    Siegel, A.R., E-mail: siegela@mcs.anl.gov [Argonne National Laboratory, Nuclear Engineering Division (United States); Argonne National Laboratory, Mathematics and Computer Science Division (United States); Smith, K., E-mail: kord@mit.edu [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering (United States); Romano, P.K., E-mail: romano7@mit.edu [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering (United States); Forget, B., E-mail: bforget@mit.edu [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering (United States); Felker, K., E-mail: felker@mcs.anl.gov [Argonne National Laboratory, Mathematics and Computer Science Division (United States)

    2013-02-15

    A model is developed to predict the impact of particle load imbalances on the performance of domain-decomposed Monte Carlo neutron transport algorithms. Expressions for upper bound performance “penalties” are derived in terms of simple machine characteristics, material characterizations and initial particle distributions. The hope is that these relations can be used to evaluate tradeoffs among different memory decomposition strategies in next generation Monte Carlo codes, and perhaps as a metric for triggering particle redistribution in production codes.

  5. An upgraded version of the nucleon meson transport code: NMTC/JAERI97

    Energy Technology Data Exchange (ETDEWEB)

    Takada, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yoshizawa, Nobuaki; Kosako, Kazuaki; Ishibashi, Kenji

    1998-02-01

    The nucleon-meson transport code NMTC/JAERI is upgraded to NMTC/JAERI97 which has new features not only in physics model and nuclear data but also in computational procedure. NMTC/JAERI97 implements the following two new physics models: an intranuclear cascade model taking account of the in-medium nuclear effects and the preequilibrium calculation model based on the exciton one. For treating the nucleon transport process more accurately, the nucleon-nucleus cross sections are revised to those derived by the systematics of Pearlstein. Moreover, the level density parameter derived by Ignatyuk is included as a new option for particle evaporation calculation. Other than those physical aspects, a new geometry package based on the Combinatorial Geometry with multi-array system and the importance sampling technique are implemented in the code. Tally function is also employed for obtaining such physical quantities as neutron energy spectra, heat deposition and nuclide yield without editing a history file. The resultant NMTC/JAERI97 is tuned to be executed on the UNIX system. This paper explains about the function, physics models and geometry model adopted in NMTC/JAERI97 and guides how to use the code. (author)

  6. Benchmarking Heavy Ion Transport Codes FLUKA, HETC-HEDS MARS15, MCNPX, and PHITS

    Energy Technology Data Exchange (ETDEWEB)

    Ronningen, Reginald Martin [Michigan State University; Remec, Igor [Oak Ridge National Laboratory; Heilbronn, Lawrence H. [University of Tennessee-Knoxville

    2013-06-07

    Powerful accelerators such as spallation neutron sources, muon-collider/neutrino facilities, and rare isotope beam facilities must be designed with the consideration that they handle the beam power reliably and safely, and they must be optimized to yield maximum performance relative to their design requirements. The simulation codes used for design purposes must produce reliable results. If not, component and facility designs can become costly, have limited lifetime and usefulness, and could even be unsafe. The objective of this proposal is to assess the performance of the currently available codes PHITS, FLUKA, MARS15, MCNPX, and HETC-HEDS that could be used for design simulations involving heavy ion transport. We plan to access their performance by performing simulations and comparing results against experimental data of benchmark quality. Quantitative knowledge of the biases and the uncertainties of the simulations is essential as this potentially impacts the safe, reliable and cost effective design of any future radioactive ion beam facility. Further benchmarking of heavy-ion transport codes was one of the actions recommended in the Report of the 2003 RIA R&D Workshop".

  7. A review of the use and potential of the GATE Monte Carlo simulation code for radiation therapy and dosimetry applications.

    Science.gov (United States)

    Sarrut, David; Bardiès, Manuel; Boussion, Nicolas; Freud, Nicolas; Jan, Sébastien; Létang, Jean-Michel; Loudos, George; Maigne, Lydia; Marcatili, Sara; Mauxion, Thibault; Papadimitroulas, Panagiotis; Perrot, Yann; Pietrzyk, Uwe; Robert, Charlotte; Schaart, Dennis R; Visvikis, Dimitris; Buvat, Irène

    2014-06-01

    In this paper, the authors' review the applicability of the open-source GATE Monte Carlo simulation platform based on the GEANT4 toolkit for radiation therapy and dosimetry applications. The many applications of GATE for state-of-the-art radiotherapy simulations are described including external beam radiotherapy, brachytherapy, intraoperative radiotherapy, hadrontherapy, molecular radiotherapy, and in vivo dose monitoring. Investigations that have been performed using GEANT4 only are also mentioned to illustrate the potential of GATE. The very practical feature of GATE making it easy to model both a treatment and an imaging acquisition within the same framework is emphasized. The computational times associated with several applications are provided to illustrate the practical feasibility of the simulations using current computing facilities.

  8. A review of the use and potential of the GATE Monte Carlo simulation code for radiation therapy and dosimetry applications

    Energy Technology Data Exchange (ETDEWEB)

    Sarrut, David, E-mail: david.sarrut@creatis.insa-lyon.fr [Université de Lyon, CREATIS, CNRS UMR5220, Inserm U1044, INSA-Lyon (France); Université Lyon 1 (France); Centre Léon Bérard (France); Bardiès, Manuel; Marcatili, Sara; Mauxion, Thibault [Inserm, UMR1037 CRCT, F-31000 Toulouse, France and Université Toulouse III-Paul Sabatier, UMR1037 CRCT, F-31000 Toulouse (France); Boussion, Nicolas [INSERM, UMR 1101, LaTIM, CHU Morvan, 29609 Brest (France); Freud, Nicolas; Létang, Jean-Michel [Université de Lyon, CREATIS, CNRS UMR5220, Inserm U1044, INSA-Lyon, Université Lyon 1, Centre Léon Bérard, 69008 Lyon (France); Jan, Sébastien [CEA/DSV/I2BM/SHFJ, Orsay 91401 (France); Loudos, George [Department of Medical Instruments Technology, Technological Educational Institute of Athens, Athens 12210 (Greece); Maigne, Lydia; Perrot, Yann [UMR 6533 CNRS/IN2P3, Université Blaise Pascal, 63171 Aubière (France); Papadimitroulas, Panagiotis [Department of Biomedical Engineering, Technological Educational Institute of Athens, 12210, Athens (Greece); Pietrzyk, Uwe [Institut für Neurowissenschaften und Medizin, Forschungszentrum Jülich GmbH, 52425 Jülich, Germany and Fachbereich für Mathematik und Naturwissenschaften, Bergische Universität Wuppertal, 42097 Wuppertal (Germany); Robert, Charlotte [IMNC, UMR 8165 CNRS, Universités Paris 7 et Paris 11, Orsay 91406 (France); and others

    2014-06-15

    In this paper, the authors' review the applicability of the open-source GATE Monte Carlo simulation platform based on the GEANT4 toolkit for radiation therapy and dosimetry applications. The many applications of GATE for state-of-the-art radiotherapy simulations are described including external beam radiotherapy, brachytherapy, intraoperative radiotherapy, hadrontherapy, molecular radiotherapy, and in vivo dose monitoring. Investigations that have been performed using GEANT4 only are also mentioned to illustrate the potential of GATE. The very practical feature of GATE making it easy to model both a treatment and an imaging acquisition within the same frameworkis emphasized. The computational times associated with several applications are provided to illustrate the practical feasibility of the simulations using current computing facilities.

  9. Physics study of microbeam radiation therapy with PSI-version of Monte Carlo code GEANT as a new computational tool

    CERN Document Server

    Stepanek, J; Laissue, J A; Lyubimova, N; Di Michiel, F; Slatkin, D N

    2000-01-01

    Microbeam radiation therapy (MRT) is a currently experimental method of radiotherapy which is mediated by an array of parallel microbeams of synchrotron-wiggler-generated X-rays. Suitably selected, nominally supralethal doses of X-rays delivered to parallel microslices of tumor-bearing tissues in rats can be either palliative or curative while causing little or no serious damage to contiguous normal tissues. Although the pathogenesis of MRT-mediated tumor regression is not understood, as in all radiotherapy such understanding will be based ultimately on our understanding of the relationships among the following three factors: (1) microdosimetry, (2) damage to normal tissues, and (3) therapeutic efficacy. Although physical microdosimetry is feasible, published information on MRT microdosimetry to date is computational. This report describes Monte Carlo-based computational MRT microdosimetry using photon and/or electron scattering and photoionization cross-section data in the 1 e V through 100 GeV range distrib...

  10. Advances in conformal radiotherapy using Monte Carlo Code to design new IMRT and IORT accelerators and interpret CT numbers

    CERN Document Server

    Wysocka-Rabin, A

    2013-01-01

    The introductory chapter of this monograph, which follows this Preface, provides an overview of radiotherapy and treatment planning. The main chapters that follow describe in detail three significant aspects of radiotherapy on which the author has focused her research efforts. Chapter 2 presents studies the author worked on at the German National Cancer Institute (DKFZ) in Heidelberg. These studies applied the Monte Carlo technique to investigate the feasibility of performing Intensity Modulated Radiotherapy (IMRT) by scanning with a narrow photon beam. This approach represents an alternative to techniques that generate beam modulation by absorption, such as MLC, individually-manufactured compensators, and special tomotherapy modulators. The technical realization of this concept required investigation of the influence of various design parameters on the final small photon beam. The photon beam to be scanned should have a diameter of approximately 5 mm at Source Surface Distance (SSD) distance, and the penumbr...

  11. Continuous-Energy Adjoint Flux and Perturbation Calculation using the Iterated Fission Probability Method in Monte Carlo Code TRIPOLI-4® and Underlying Applications

    Science.gov (United States)

    Truchet, G.; Leconte, P.; Peneliau, Y.; Santamarina, A.; Malvagi, F.

    2014-06-01

    Pile-oscillation experiments are performed in the MINERVE reactor at the CEA Cadarache to improve nuclear data accuracy. In order to precisely calculate small reactivity variations (experiments, a reference calculation need to be achieved. This calculation may be accomplished using the continuous-energy Monte Carlo code TRIPOLI-4® by using the eigenvalue difference method. This "direct" method has shown limitations in the evaluation of very small reactivity effects because it needs to reach a very small variance associated to the reactivity in both states. To answer this problem, it has been decided to implement the exact perturbation theory in TRIPOLI-4® and, consequently, to calculate a continuous-energy adjoint flux. The Iterated Fission Probability (IFP) method was chosen because it has shown great results in some other Monte Carlo codes. The IFP method uses a forward calculation to compute the adjoint flux, and consequently, it does not rely on complex code modifications but on the physical definition of the adjoint flux as a phase-space neutron importance. In the first part of this paper, the IFP method implemented in TRIPOLI-4® is described. To illustrate the effciency of the method, several adjoint fluxes are calculated and compared with their equivalent obtained by the deterministic code APOLLO-2. The new implementation can calculate angular adjoint flux. In the second part, a procedure to carry out an exact perturbation calculation is described. A single cell benchmark has been used to test the accuracy of the method, compared with the "direct" estimation of the perturbation. Once again the method based on the IFP shows good agreement for a calculation time far more inferior to the "direct" method. The main advantage of the method is that the relative accuracy of the reactivity variation does not depend on the magnitude of the variation itself, which allows us to calculate very small reactivity perturbations with high precision. Other applications of

  12. Lower Hybrid Current Drive and Heating for the National Transport Code Collaboration

    Science.gov (United States)

    Ignat, D. W.; Jardin, S. C.; McCune, D. C.; Valeo, E. J.

    2000-10-01

    The Lower hybrid Simulation Code LSC was originally written as a subroutine to the Toroidal Simulation Code TSC (Jardin, Pomphrey, Kessel, et al) and subsequently ported to a subroutine of TRANSP. Modifications to simplify the use of the LSC both as a callable module, and also independently of larger transport codes, and improve the documentation have been undertaken with the goal of installing LSC in the NTCC library. The physical model, which includes ray tracing from a Brambilla spectrum, 1D Fokker-Planck development of the electron distribution, the Karney-Fisch treatment of the electric field, heuristic diffusion of current and power and wall scattering, has not been changed. The computational approach is to suppress or remove from the control of the user numerical parameters such as step size and number of iterations while changing some code to be extremely stable in varied conditions. Essential graphics are now output as gnuplot commands and data for off-line post processing, but the original outputs to sglib are retained as an option. Examples of output are shown.

  13. Parallelization of a three-dimensional whole core transport code DeCART

    Energy Technology Data Exchange (ETDEWEB)

    Jin Young, Cho; Han Gyu, Joo; Ha Yong, Kim; Moon-Hee, Chang [Korea Atomic Energy Research Institute, Yuseong-gu, Daejon (Korea, Republic of)

    2003-07-01

    Parallelization of the DeCART (deterministic core analysis based on ray tracing) code is presented that reduces the computational burden of the tremendous computing time and memory required in three-dimensional whole core transport calculations. The parallelization employs the concept of MPI grouping and the MPI/OpenMP mixed scheme as well. Since most of the computing time and memory are used in MOC (method of characteristics) and the multi-group CMFD (coarse mesh finite difference) calculation in DeCART, variables and subroutines related to these two modules are the primary targets for parallelization. Specifically, the ray tracing module was parallelized using a planar domain decomposition scheme and an angular domain decomposition scheme. The parallel performance of the DeCART code is evaluated by solving a rodded variation of the C5G7MOX three dimensional benchmark problem and a simplified three-dimensional SMART PWR core problem. In C5G7MOX problem with 24 CPUs, a speedup of maximum 21 is obtained on an IBM Regatta machine and 22 on a LINUX Cluster in the MOC kernel, which indicates good parallel performance of the DeCART code. In the simplified SMART problem, the memory requirement of about 11 GBytes in the single processor cases reduces to 940 Mbytes with 24 processors, which means that the DeCART code can now solve large core problems with affordable LINUX clusters. (authors)

  14. PRESTO-II: a low-level waste environmental transport and risk assessment code

    Energy Technology Data Exchange (ETDEWEB)

    Fields, D.E.; Emerson, C.J.; Chester, R.O.; Little, C.A.; Hiromoto, G.

    1986-04-01

    PRESTO-II (Prediction of Radiation Effects from Shallow Trench Operations) is a computer code designed for the evaluation of possible health effects from shallow-land and, waste-disposal trenches. The model is intended to serve as a non-site-specific screening model for assessing radionuclide transport, ensuing exposure, and health impacts to a static local population for a 1000-year period following the end of disposal operations. Human exposure scenarios considered include normal releases (including leaching and operational spillage), human intrusion, and limited site farming or reclamation. Pathways and processes of transit from the trench to an individual or population include ground-water transport, overland flow, erosion, surface water dilution, suspension, atmospheric transport, deposition, inhalation, external exposure, and ingestion of contaminated beef, milk, crops, and water. Both population doses and individual doses, as well as doses to the intruder and farmer, may be calculated. Cumulative health effects in terms of cancer deaths are calculated for the population over the 1000-year period using a life-table approach. Data are included for three example sites: Barnwell, South Carolina; Beatty, Nevada; and West Valley, New York. A code listing and example input for each of the three sites are included in the appendices to this report.

  15. Path-integral Monte Carlo simulations for electronic dynamics on molecular chains. II. Transport across impurities

    Science.gov (United States)

    Mühlbacher, Lothar; Ankerhold, Joachim

    2005-05-01

    Electron transfer (ET) across molecular chains including an impurity is studied based on a recently improved real-time path-integral Monte Carlo (PIMC) approach [L. Mühlbacher, J. Ankerhold, and C. Escher, J. Chem. Phys. 121 12696 (2004)]. The reduced electronic dynamics is studied for various bridge lengths and defect site energies. By determining intersite hopping rates from PIMC simulations up to moderate times, the relaxation process in the extreme long-time limit is captured within a sequential transfer model. The total transfer rate is extracted and shown to be enhanced for certain defect site energies. Superexchange turns out to be relevant for extreme gap energies only and then gives rise to different dynamical signatures for high- and low-lying defects. Further, it is revealed that the entire bridge compound approaches a steady state on a much shorter time scale than that related to the total transfer. This allows for a simplified description of ET along donor-bridge-acceptor systems in the long-time range.

  16. OpenGeoSys-GEMS: Hybrid parallelization of a reactive transport code with MPI and threads

    Science.gov (United States)

    Kosakowski, G.; Kulik, D. A.; Shao, H.

    2012-04-01

    OpenGeoSys-GEMS is a generic purpose reactive transport code based on the operator splitting approach. The code couples the Finite-Element groundwater flow and multi-species transport modules of the OpenGeoSys (OGS) project (http://www.ufz.de/index.php?en=18345) with the GEM-Selektor research package to model thermodynamic equilibrium of aquatic (geo)chemical systems utilizing the Gibbs Energy Minimization approach (http://gems.web.psi.ch/). The combination of OGS and the GEM-Selektor kernel (GEMS3K) is highly flexible due to the object-oriented modular code structures and the well defined (memory based) data exchange modules. Like other reactive transport codes, the practical applicability of OGS-GEMS is often hampered by the long calculation time and large memory requirements. • For realistic geochemical systems which might include dozens of mineral phases and several (non-ideal) solid solutions the time needed to solve the chemical system with GEMS3K may increase exceptionally. • The codes are coupled in a sequential non-iterative loop. In order to keep the accuracy, the time step size is restricted. In combination with a fine spatial discretization the time step size may become very small which increases calculation times drastically even for small 1D problems. • The current version of OGS is not optimized for memory use and the MPI version of OGS does not distribute data between nodes. Even for moderately small 2D problems the number of MPI processes that fit into memory of up-to-date workstations or HPC hardware is limited. One strategy to overcome the above mentioned restrictions of OGS-GEMS is to parallelize the coupled code. For OGS a parallelized version already exists. It is based on a domain decomposition method implemented with MPI and provides a parallel solver for fluid and mass transport processes. In the coupled code, after solving fluid flow and solute transport, geochemical calculations are done in form of a central loop over all finite

  17. Robust and accurate transient light transport decomposition via convolutional sparse coding.

    Science.gov (United States)

    Hu, Xuemei; Deng, Yue; Lin, Xing; Suo, Jinli; Dai, Qionghai; Barsi, Christopher; Raskar, Ramesh

    2014-06-01

    Ultrafast sources and detectors have been used to record the time-resolved scattering of light propagating through macroscopic scenes. In the context of computational imaging, decomposition of this transient light transport (TLT) is useful for applications, such as characterizing materials, imaging through diffuser layers, and relighting scenes dynamically. Here, we demonstrate a method of convolutional sparse coding to decompose TLT into direct reflections, inter-reflections, and subsurface scattering. The method relies on the sparsity composition of the time-resolved kernel. We show that it is robust and accurate to noise during the acquisition process.

  18. Polarization correction for ionization loss in a galactic cosmic ray transport code (HZETRN)

    Science.gov (United States)

    Shinn, Judy L.; Farhat, Hamidullah; Badavi, Francis F.; Wilson, John W.

    1993-03-01

    An approximate polarization correction for ionization loss suggested by Sternheimer has been implemented in the galactic cosmic ray transport code (HZETRN) developed at the Langley Research Center. Sample calculations made for the aluminum shield and liquid hydrogen shield show no more than a plus or minus 2 percent change in the linear energy transfer (LET) distribution for flux compared with those without polarization correction. This very small change is expected because the effect of polarization correction on the reduction in stopping power of ions with energies above 2 GeV/amu is suppressed by the decrease in galactic cosmic ray ion flux at such high energies.

  19. SHIELD-HIT12A - a Monte Carlo particle transport program for ion therapy research

    DEFF Research Database (Denmark)

    Bassler, Niels; Hansen, David Christoffer; Lühr, Armin;

    2014-01-01

    -HIT to a heavy ion dose optimization algorithm to provide MC-optimized treatment plans that include radiobiology. Methods: SHIELD-HIT12A is written in FORTRAN and carefully retains platform independence. A powerful scoring engine is implemented scoring relevant quantities such as dose and track-average LET....... We experienced that new users quickly learn to use SHIELD-HIT12A and setup new geometries. Contrary to previous versions of SHIELD-HIT, the 12A distribution comes along with easy-to-use example files and an English manual. A new implementation of Vavilov straggling resulted in a massive reduction...... of computation time. Scheduled for later release are CT import and photon-electron transport. Conclusions: SHIELD-HIT12A is an interesting alternative ion transport engine. Apart from being a flexible particle therapy research tool, it can also serve as a back end for a MC ion treatment planning system. More...

  20. Internal dosimetry with the Monte Carlo code GATE: validation using the ICRP/ICRU female reference computational model

    Science.gov (United States)

    Villoing, Daphnée; Marcatili, Sara; Garcia, Marie-Paule; Bardiès, Manuel

    2017-03-01

    The purpose of this work was to validate GATE-based clinical scale absorbed dose calculations in nuclear medicine dosimetry. GATE (version 6.2) and MCNPX (version 2.7.a) were used to derive dosimetric parameters (absorbed fractions, specific absorbed fractions and S-values) for the reference female computational model proposed by the International Commission on Radiological Protection in ICRP report 110. Monoenergetic photons and electrons (from 50 keV to 2 MeV) and four isotopes currently used in nuclear medicine (fluorine-18, lutetium-177, iodine-131 and yttrium-90) were investigated. Absorbed fractions, specific absorbed fractions and S-values were generated with GATE and MCNPX for 12 regions of interest in the ICRP 110 female computational model, thereby leading to 144 source/target pair configurations. Relative differences between GATE and MCNPX obtained in specific configurations (self-irradiation or cross-irradiation) are presented. Relative differences in absorbed fractions, specific absorbed fractions or S-values are below 10%, and in most cases less than 5%. Dosimetric results generated with GATE for the 12 volumes of interest are available as supplemental data. GATE can be safely used for radiopharmaceutical dosimetry at the clinical scale. This makes GATE a viable option for Monte Carlo modelling of both imaging and absorbed dose in nuclear medicine.

  1. Benchmarking PARTISN with Analog Monte Carlo: Moments of the Neutron Number and the Cumulative Fission Number Probability Distributions

    Energy Technology Data Exchange (ETDEWEB)

    O' Rourke, Patrick Francis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-10-27

    The purpose of this report is to provide the reader with an understanding of how a Monte Carlo neutron transport code was written, developed, and evolved to calculate the probability distribution functions (PDFs) and their moments for the neutron number at a final time as well as the cumulative fission number, along with introducing several basic Monte Carlo concepts.

  2. 3D Monte-Carlo transport calculations of whole slab reactor cores: validation of deterministic neutronic calculation routes

    Energy Technology Data Exchange (ETDEWEB)

    Palau, J.M. [CEA Cadarache, Service de Physique des Reacteurs et du Cycle, Lab. de Projets Nucleaires, 13 - Saint-Paul-lez-Durance (France)

    2005-07-01

    This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U{sup 235}, U{sup 238}, Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)

  3. AEOLUS: A MARKOV CHAIN MONTE CARLO CODE FOR MAPPING ULTRACOOL ATMOSPHERES. AN APPLICATION ON JUPITER AND BROWN DWARF HST LIGHT CURVES

    Energy Technology Data Exchange (ETDEWEB)

    Karalidi, Theodora; Apai, Dániel; Schneider, Glenn; Hanson, Jake R. [Steward Observatory, Department of Astronomy, University of Arizona, 933 N. Cherry Avenue, Tucson, AZ 85721 (United States); Pasachoff, Jay M., E-mail: tkaralidi@email.arizona.edu [Hopkins Observatory, Williams College, 33 Lab Campus Drive, Williamstown, MA 01267 (United States)

    2015-11-20

    Deducing the cloud cover and its temporal evolution from the observed planetary spectra and phase curves can give us major insight into the atmospheric dynamics. In this paper, we present Aeolus, a Markov chain Monte Carlo code that maps the structure of brown dwarf and other ultracool atmospheres. We validated Aeolus on a set of unique Jupiter Hubble Space Telescope (HST) light curves. Aeolus accurately retrieves the properties of the major features of the Jovian atmosphere, such as the Great Red Spot and a major 5 μm hot spot. Aeolus is the first mapping code validated on actual observations of a giant planet over a full rotational period. For this study, we applied Aeolus to J- and H-band HST light curves of 2MASS J21392676+0220226 and 2MASS J0136565+093347. Aeolus retrieves three spots at the top of the atmosphere (per observational wavelength) of these two brown dwarfs, with a surface coverage of 21% ± 3% and 20.3% ± 1.5%, respectively. The Jupiter HST light curves will be publicly available via ADS/VIZIR.

  4. Investigation of behavior of scintillator detector of Alborz observatory array using Monte Carlo method with Geant4 code

    Directory of Open Access Journals (Sweden)

    M. Abbasian Motlagh

    2014-04-01

    Full Text Available For their appropriate temporal resolution, scintillator detectors are used in the Alborz observatory. In this work, the behavior of the scintillation detectors for the passage of electrons with different energies and directions were studied using the simulation code GEANT4. Pulse shapes of scintillation light, and such characteristics as the total number of photons, the rise time and the falling time for the optical pulses were computed for the passage of electrons with energies of 10, 100 and 1000 MeV. Variations of the characteristics of optical pulse of scintillation with incident angle and the location of electrons were also investigated

  5. Understanding transport simulations of heavy-ion collisions at 100 and 400 AMeV: Comparison of heavy ion transport codes under controlled conditions

    CERN Document Server

    Xu, Jun; Tsang, ManYee Betty; Wolter, Hermann; Zhang, Ying-Xun; Aichelin, Joerg; Colonna, Maria; Cozma, Dan; Danielewicz, Pawel; Feng, Zhao-Qing; Fevre, Arnaud Le; Gaitanos, Theodoros; Hartnack, Christoph; Kim, Kyungil; Kim, Youngman; Ko, Che-Ming; Li, Bao-An; Li, Qing-Feng; Li, Zhu-Xia; Napolitani, Paolo; Ono, Akira; Papa, Massimo; Song, Taesoo; Su, Jun; Tian, Jun-Long; Wang, Ning; Wang, Yong-Jia; Weil, Janus; Xie, Wen-Jie; Zhang, Feng-Shou; Zhang, Guo-Qiang

    2016-01-01

    Transport simulations are very valuable for extracting physics information from heavy-ion collision experiments. With the emergence of many different transport codes in recent years, it becomes important to estimate their robustness in extracting physics information from experiments. We report on the results of a transport code comparison project. 18 commonly used transport codes were included in this comparison: 9 Boltzmann-Uehling-Uhlenbeck-type codes and 9 Quantum-Molecular-Dynamics-type codes. These codes have been required to simulate Au+Au collisions using the same physics input for mean fields and for in-medium nucleon-nucleon cross sections, as well as the same initialization set-up, the impact parameter, and other calculational parameters at 100 and 400 AMeV incident energy. Among the codes we compare one-body observables such as rapidity and transverse flow distributions. We also monitor non-observables such as the initialization of the internal states of colliding nuclei and their stability, the co...

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

    Energy Technology Data Exchange (ETDEWEB)

    Mitchell, Dean J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Contraband Detection; Thoreson, Gregory G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Contraband Detection; Horne, Steven M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Contraband Detection

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mitchell, Dean J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Contraband Detection; Thoreson, Gregory G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Contraband Detection; Horne, Steven M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Contraband Detection

    2014-01-01

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

  8. Construction of a computational exposure model for dosimetric calculations using the EGS4 Monte Carlo code and voxel phantoms; Construcao de um modelo computacional de exposicao para calculos dosimetricos utilizando o codigo Monte Carlo EGS4 e fantomas de voxels

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Jose Wilson

    2004-07-15

    The MAX phantom has been developed from existing segmented images of a male adult body, in order to achieve a representation as close as possible to the anatomical properties of the reference adult male specified by the ICRP. In computational dosimetry, MAX can simulate the geometry of a human body under exposure to ionizing radiations, internal or external, with the objective of calculating the equivalent dose in organs and tissues for occupational, medical or environmental purposes of the radiation protection. This study presents a methodology used to build a new computational exposure model MAX/EGS4: the geometric construction of the phantom; the development of the algorithm of one-directional, divergent, and isotropic radioactive sources; new methods for calculating the equivalent dose in the red bone marrow and in the skin, and the coupling of the MAX phantom with the EGS4 Monte Carlo code. Finally, some results of radiation protection, in the form of conversion coefficients between equivalent dose (or effective dose) and free air-kerma for external photon irradiation are presented and discussed. Comparing the results presented with similar data from other human phantoms it is possible to conclude that the coupling MAX/EGS4 is satisfactory for the calculation of the equivalent dose in radiation protection. (author)

  9. Validation of the coupling of mesh models to GEANT4 Monte Carlo code for simulation of internal sources of photons; Validacao do acoplamento de modelos mesh ao codigo Monte Carlo GEANT4 para simulacao de fontes de fotons internas

    Energy Technology Data Exchange (ETDEWEB)

    Caribe, Paulo Rauli Rafeson Vasconcelos, E-mail: raulycaribe@hotmail.com [Universidade Federal Rural de Pernambuco (UFRPE), Recife, PE (Brazil). Fac. de Fisica; Cassola, Vagner Ferreira; Kramer, Richard; Khoury, Helen Jamil [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear

    2013-07-01

    The use of three-dimensional models described by polygonal meshes in numerical dosimetry enables more accurate modeling of complex objects than the use of simple solid. The objectives of this work were validate the coupling of mesh models to the Monte Carlo code GEANT4 and evaluate the influence of the number of vertices in the simulations to obtain absorbed fractions of energy (AFEs). Validation of the coupling was performed to internal sources of photons with energies between 10 keV and 1 MeV for spherical geometries described by the GEANT4 and three-dimensional models with different number of vertices and triangular or quadrilateral faces modeled using Blender program. As a result it was found that there were no significant differences between AFEs for objects described by mesh models and objects described using solid volumes of GEANT4. Since that maintained the shape and the volume the decrease in the number of vertices to describe an object does not influence so meant dosimetric data, but significantly decreases the time required to achieve the dosimetric calculations, especially for energies less than 100 keV.

  10. 76 FR 2744 - Disclosure of Code-Share Service by Air Carriers and Sellers of Air Transportation

    Science.gov (United States)

    2011-01-14

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF TRANSPORTATION... agents must inform the consumer of the code-share service ``before booking transportation'' and state... global distribution systems, which may be assisting travel agents to establish airline ticket sales...

  11. Verification of high-energy transport codes on the basis of activation data

    CERN Document Server

    Titarenko, Yu E; Butko, M A; Dikarev, D V; Florya, S N; Pavlov, K V; Titarenko, A Yu; Tikhonov, R S; Zhivun, V M; Ignatyuk, A V; Mashnik, S G; Boudard, A; Leray, S; David, J -C; Cugnon, J; Mancusi, D; Yariv, Y; Kumawat, H; Nishihara, K; Matsuda, N; Mank, G; Gudowski, W

    2011-01-01

    Nuclide production cross sections measured at ITEP for the targets of nat-Cr, 56-Fe, nat-Ni, 93-Nb, 181-Ta, nat-W, nat-Pb, 209-Bi irradiated by protons with energies from 40 to 2600 MeV were used to estimate the predictive accuracy of several popular high-energy transport codes. A general agreement of the ITEP data with the data obtained by other groups, including the numerous GSI data measured by the inverse kinematics method was found. Simulations of the measured data were performed with the MCNPX (Bertini and ISABEL options), CEM03.02, INCL4.2+ABLA, INCL4.5+ABLA07, PHITS, and CASCADE.07 codes. Deviation factors between the calculated and experimental cross sections have been estimated for each target and for the whole energy range covered by our measurements. Two-dimensional diagrams of deviation factor values were produced for estimating the predictive power of every code for intermediate, not measured masses of nuclei-targets and bombarding energies of protons. Further improvements of all tested here cod...

  12. Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system

    Energy Technology Data Exchange (ETDEWEB)

    Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi

    1998-01-01

    In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)

  13. Summary report for ITER task - D10: Update and implementation of neutron transport and activation codes and processed libraries

    Energy Technology Data Exchange (ETDEWEB)

    Attaya, H.

    1995-01-01

    The primary goal of this task is to provide the capabilities in the activation code RACC, to treat pulsed operation modes. In addition, it is required that the code utilizes the same spatial mesh and geometrical models as employed in the one or multidimensional neutron transport codes used in ITER design. This would ensure the use of the same neutron flux generated by those codes to calculate the different activation parameters. It is also required to have the capabilities for generating graphical outputs for the calculated activation parameters.

  14. Kinetic Monte Carlo of transport processes in Al/AlOx/Au-layers: Impact of defects

    Science.gov (United States)

    Weiler, Benedikt; Haeberle, Tobias; Gagliardi, Alessio; Lugli, Paolo

    2016-09-01

    Ultrathin films of alumina were investigated by a compact kMC-model. Experimental jV-curves from Al/AlOx/Au-junctions with plasma- and thermal-grown AlOx were fitted by simulated ones. We found dominant defects at 2.3-2.5 eV below CBM for AlOx with an effective mass mox ∗= 0.35 m0 and a barrier EB ,A l /A l O x≈2.8 eV in agreement with literature. The parameterization is extended to varying defect levels, defect densities, injection barriers, effective masses and the thickness of AlOx. Thus, dominant charge transport processes and implications on the relevance of defects are derived and AlOx parameters are specified which are detrimental for the operation of devices.

  15. Kinetic Monte Carlo of transport processes in Al/AlOx/Au-layers: Impact of defects

    Directory of Open Access Journals (Sweden)

    Benedikt Weiler

    2016-09-01

    Full Text Available Ultrathin films of alumina were investigated by a compact kMC-model. Experimental jV-curves from Al/AlOx/Au-junctions with plasma- and thermal-grown AlOx were fitted by simulated ones. We found dominant defects at 2.3-2.5 eV below CBM for AlOx with an effective mass mox∗=0.35 m0 and a barrier EB,Al/AlOx≈2.8 eV in agreement with literature. The parameterization is extended to varying defect levels, defect densities, injection barriers, effective masses and the thickness of AlOx. Thus, dominant charge transport processes and implications on the relevance of defects are derived and AlOx parameters are specified which are detrimental for the operation of devices.

  16. Angular Distribution of Particles Emerging from a Diffusive Region and its Implications for the Fleck-Canfield Random Walk Algorithm for Implicit Monte Carlo Radiation Transport

    CERN Document Server

    Cooper, M A

    2000-01-01

    We present various approximations for the angular distribution of particles emerging from an optically thick, purely isotropically scattering region into a vacuum. Our motivation is to use such a distribution for the Fleck-Canfield random walk method [1] for implicit Monte Carlo (IMC) [2] radiation transport problems. We demonstrate that the cosine distribution recommended in the original random walk paper [1] is a poor approximation to the angular distribution predicted by transport theory. Then we examine other approximations that more closely match the transport angular distribution.

  17. Design of Light Multi-layered Shields for Use in Diagnostic Radiology and Nuclear Medicine via MCNP5 Monte Carlo Code

    Directory of Open Access Journals (Sweden)

    Mehdi Zehtabian

    2015-09-01

    Full Text Available Introduction Lead-based shields are the most widely used attenuators in X-ray and gamma ray fields. The heavy weight, toxicity and corrosion of lead have led researchers towards the development of non-lead shields. Materials and Methods The purpose of this study was to design multi-layered shields for protection against X-rays and gamma rays in diagnostic radiology and nuclear medicine. In this study, cubic slabs composed of several materials with high atomic numbers, i.e., lead, barium, bismuth, gadolinium, tin and tungsten, were simulated, using MCNP5 Monte Carlo code. Cubic slabs (30×30×0.05 cm3 were simulated at a 50 cm distance from the point photon source. The X-ray spectra of 80 kVp and 120 kVp were obtained, using IPEM Report 78. The photon flux following the use of each shield was obtained inside cubic tally cells (1×1×0.5 cm3 at a 5 cm distance from the shields. The photon attenuation properties of multi-layered shields (i.e., two, three, four and five layers, composed of non-lead radiation materials, were also obtained via Monte Carlo simulations. Results Among different shield designs proposed in this study, the three-layered shield, composed of tungsten, bismuth and gadolinium, showed the most significant attenuation properties in radiology, with acceptable shielding at 140 keV energy in nuclear medicine. Conclusion According to the results, materials with k-edges equal to energies common to diagnostic radiology can be proper substitutes for lead shields.

  18. Core 2D. A code for non-isothermal water flow and reactive solute transport. Users manual version 2

    Energy Technology Data Exchange (ETDEWEB)

    Samper, J.; Juncosa, R.; Delgado, J.; Montenegro, L. [Universidad de A Coruna (Spain)

    2000-07-01

    Understanding natural groundwater quality patterns, quantifying groundwater pollution and assessing the effects of waste disposal, require modeling tools accounting for water flow, and transport of heat and dissolved species as well as their complex interactions with solid and gases phases. This report contains the users manual of CORE ''2D Version V.2.0, a COde for modeling water flow (saturated and unsaturated), heat transport and multicomponent Reactive solute transport under both local chemical equilibrium and kinetic conditions. it is an updated and improved version of CORE-LE-2D V0 (Samper et al., 1988) which in turns is an extended version of TRANQUI, a previous reactive transport code (ENRESA, 1995). All these codes were developed within the context of Research Projects funded by ENRESA and the European Commission. (Author)

  19. Monte Carlo simulation of radiation heat transfer in arrays of fixed discrete surfaces using cell-to-cell photon transport

    Energy Technology Data Exchange (ETDEWEB)

    Drost, M.K. [Pacific Northwest Lab., Richland, WA (United States); Welty, J.R. [Oregon State Univ., Corvallis, OR (United States)

    1992-08-01

    Radiation heat transfer in an array of fixed discrete surfaces is an important problem that is particularly difficult to analyze because of the nonhomogeneous and anisotropic optical properties involved. This article presents an efficient Monte Carlo method for evaluating radiation heat transfer in arrays of fixed discrete surfaces. This Monte Carlo model has been optimized to take advantage of the regular arrangement of surfaces often encountered in these arrays. Monte Carlo model predictions have been compared with analytical and experimental results.

  20. Elucidating the electron transport in semiconductors via Monte Carlo simulations: an inquiry-driven learning path for engineering undergraduates

    Science.gov (United States)

    Persano Adorno, Dominique; Pizzolato, Nicola; Fazio, Claudio

    2015-09-01

    Within the context of higher education for science or engineering undergraduates, we present an inquiry-driven learning path aimed at developing a more meaningful conceptual understanding of the electron dynamics in semiconductors in the presence of applied electric fields. The electron transport in a nondegenerate n-type indium phosphide bulk semiconductor is modelled using a multivalley Monte Carlo approach. The main characteristics of the electron dynamics are explored under different values of the driving electric field, lattice temperature and impurity density. Simulation results are presented by following a question-driven path of exploration, starting from the validation of the model and moving up to reasoned inquiries about the observed characteristics of electron dynamics. Our inquiry-driven learning path, based on numerical simulations, represents a viable example of how to integrate a traditional lecture-based teaching approach with effective learning strategies, providing science or engineering undergraduates with practical opportunities to enhance their comprehension of the physics governing the electron dynamics in semiconductors. Finally, we present a general discussion about the advantages and disadvantages of using an inquiry-based teaching approach within a learning environment based on semiconductor simulations.

  1. Extremity dosimetry problems during the handling of radionuclides syringes in nuclear medicine: A Monte Carlo radiation transport simplified approach

    Energy Technology Data Exchange (ETDEWEB)

    Mariotti, F., E-mail: francesca.mariotti@bologna.enea.i [ENEA-BAS-ION IRP Radiation Protection Institute, Via dei Colli 16, 40136, Bologna (Italy); Gualdrini, G. [ENEA-BAS-ION IRP Radiation Protection Institute, Via dei Colli 16, 40136, Bologna (Italy)

    2011-04-15

    The ORAMED (Optimization of RAdiation protection for MEDical staff) Working Tasks (WP4) is addressed at evaluating extremity doses (and dose distributions across the hands) of medical staff working in nuclear medicine departments, to study the influence of protective devices such as syringe and vial shields, to improve such devices when possible and to propose 'levels of reference doses' for each standard nuclear medicine procedure. In particular task 4 is concerned with the study of the extremity dosimetry for the hand of operators devoted to the preparation and administration stages of the usage, for example, of {sup 99m}Tc, {sup 18}F and {sup 90}Y (Zevalin) radionuclides. The aim of this report consists in the study of photon-electron equilibrium conditions at 0.07 mm in the skin to justify a simplified 'kerma approximation' approach in the planned complex Monte Carlo voxel hand modeling. Furthermore a detailed investigation on primary electron and secondary bremsstrahlung photon transport from {sup 90}Y to speed up the calculations was performed. The results obtained in the simplified investigated conditions could be of help for the production calculations, introducing, if necessary, suited correction factors applicable to the complex condition results.

  2. Calculation of electron dose to target cells in a complex environment by Monte Carlo code ''CELLDOSE''

    Energy Technology Data Exchange (ETDEWEB)

    Hindie, Elif; Moretti, Jean-Luc [Hopital Saint-Louis, Service de Medecine Nucleaire, Paris (France)]|[Universite Paris 7, Imagerie Moleculaire Diagnostique et Ciblage Therapeutique, Paris (France); Champion, Christophe [Universite Paul Verlaine, Laboratoire de Physique Moleculaire et des Collisions, Metz Institut de Physique, Metz (France); Zanotti-Fregonara, Paolo; Ravasi, Laura [Commissariat a l' Energie Atomique, DSV/I2BM/SHFJ/LIME, Orsay (France); Rubello, Domenico [Instituto Oncologico Veneto (IOV) - IRCCS, Department of Nuclear Medicine - PET Centre, Rovigo (Italy); Colas-Linhart, Nicole [Faculte de Medecine Xavier Bichat, Laboratoire de Biophysique, Paris (France)

    2009-01-15

    We used the Monte Carlo code ''CELLDOSE'' to assess the dose received by specific target cells from electron emissions in a complex environment. {sup 131}I in a simulated thyroid was used as a model. Thyroid follicles were represented by 170{mu}m diameter spherical units made of a lumen of 150{mu}m diameter containing colloidal matter and a peripheral layer of 10{mu}m thick thyroid cells. Neighbouring follicles are 4{mu}m apart. {sup 131}I was assumed to be homogeneously distributed in the lumen and absent in cells. We firstly assessed electron dose distribution in a single follicle. Then, we expanded the simulation by progressively adding neighbouring layers of follicles, so to reassess the electron dose to this single follicle implemented with the contribution of the added layers. Electron dose gradient around a point source showed that the {sup 131}I electron dose is close to zero after 2,100{mu}m. Therefore, we studied all contributions to the central follicle deriving from follicles within 12 orders of neighbourhood (15,624 follicles surrounding the central follicle). The dose to colloid of the single follicle was twice as high as the dose to thyroid cells. Even when all neighbours were taken into account, the dose in the central follicle remained heterogeneous. For a 1-Gy average dose to tissue, the dose to colloidal matter was 1.168 Gy, the dose to thyroid cells was 0.982 Gy, and the dose to the inter-follicular tissue was 0.895 Gy. Analysis of the different contributions to thyroid cell dose showed that 17.3% of the dose derived from the colloidal matter of their own follicle, while the remaining 82.7% was delivered by the surrounding follicles. On the basis of these data, it is shown that when different follicles contain different concentrations of {sup 131}I, the impact in terms of cell dose heterogeneity can be important. By means of {sup 131}I in the thyroid as a theoretical model, we showed how a Monte Carlo code can be used to map

  3. Monte Carlo simulation of neutron scattering instruments

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, P.A.; Daemen, L.L.; Hjelm, R.P. Jr.

    1998-12-01

    A code package consisting of the Monte Carlo Library MCLIB, the executing code MC{_}RUN, the web application MC{_}Web, and various ancillary codes is proposed as an open standard for simulation of neutron scattering instruments. The architecture of the package includes structures to define surfaces, regions, and optical elements contained in regions. A particle is defined by its vector position and velocity, its time of flight, its mass and charge, and a polarization vector. The MC{_}RUN code handles neutron transport and bookkeeping, while the action on the neutron within any region is computed using algorithms that may be deterministic, probabilistic, or a combination. Complete versatility is possible because the existing library may be supplemented by any procedures a user is able to code. Some examples are shown.

  4. Guidelines for selecting codes for ground-water transport modeling of low-level waste burial sites. Executive summary

    Energy Technology Data Exchange (ETDEWEB)

    Simmons, C.S.; Cole, C.R.

    1985-05-01

    This document was written to provide guidance to managers and site operators on how ground-water transport codes should be selected for assessing burial site performance. There is a need for a formal approach to selecting appropriate codes from the multitude of potentially useful ground-water transport codes that are currently available. Code selection is a problem that requires more than merely considering mathematical equation-solving methods. These guidelines are very general and flexible and are also meant for developing systems simulation models to be used to assess the environmental safety of low-level waste burial facilities. Code selection is only a single aspect of the overall objective of developing a systems simulation model for a burial site. The guidance given here is mainly directed toward applications-oriented users, but managers and site operators need to be familiar with this information to direct the development of scientifically credible and defensible transport assessment models. Some specific advice for managers and site operators on how to direct a modeling exercise is based on the following five steps: identify specific questions and study objectives; establish costs and schedules for achieving answers; enlist the aid of professional model applications group; decide on approach with applications group and guide code selection; and facilitate the availability of site-specific data. These five steps for managers/site operators are discussed in detail following an explanation of the nine systems model development steps, which are presented first to clarify what code selection entails.

  5. Comparison of depth-dose distributions of proton therapeutic beams calculated by means of logical detectors and ionization chamber modeled in Monte Carlo codes

    Science.gov (United States)

    Pietrzak, Robert; Konefał, Adam; Sokół, Maria; Orlef, Andrzej

    2016-08-01

    The success of proton therapy depends strongly on the precision of treatment planning. Dose distribution in biological tissue may be obtained from Monte Carlo simulations using various scientific codes making it possible to perform very accurate calculations. However, there are many factors affecting the accuracy of modeling. One of them is a structure of objects called bins registering a dose. In this work the influence of bin structure on the dose distributions was examined. The MCNPX code calculations of Bragg curve for the 60 MeV proton beam were done in two ways: using simple logical detectors being the volumes determined in water, and using a precise model of ionization chamber used in clinical dosimetry. The results of the simulations were verified experimentally in the water phantom with Marcus ionization chamber. The average local dose difference between the measured relative doses in the water phantom and those calculated by means of the logical detectors was 1.4% at first 25 mm, whereas in the full depth range this difference was 1.6% for the maximum uncertainty in the calculations less than 2.4% and for the maximum measuring error of 1%. In case of the relative doses calculated with the use of the ionization chamber model this average difference was somewhat greater, being 2.3% at depths up to 25 mm and 2.4% in the full range of depths for the maximum uncertainty in the calculations of 3%. In the dose calculations the ionization chamber model does not offer any additional advantages over the logical detectors. The results provided by both models are similar and in good agreement with the measurements, however, the logical detector approach is a more time-effective method.

  6. Parallel processing method for two-dimensional Sn transport code DOT3.5

    Energy Technology Data Exchange (ETDEWEB)

    Uematsu, Mikio [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    1998-03-01

    A parallel processing method for the two-dimensional Sn transport code DOT3.5 has been developed to achieve drastic reduction of computation time. In the proposed method, parallelization is made with angular domain decomposition and/or space domain decomposition. Calculational speedup for parallel processing by angular domain decomposition is achieved by minimizing frequency of communications between processing elements. As for parallel processing by space domain decomposition, two-step rescaling method consisting of segmentwise rescaling and the ordinary pointwise rescaling have been developed to accelerate convergence, which will otherwise be degraded because of discontinuity at the segment boundaries. The developed method was examined with a Sun workstation using the PVM message-passing library, and sufficient speedup was observed. (author)

  7. Radiation Transport Calculations and Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Fasso, Alberto; /SLAC; Ferrari, A.; /CERN

    2011-06-30

    This article is an introduction to the Monte Carlo method as used in particle transport. After a description at an elementary level of the mathematical basis of the method, the Boltzmann equation and its physical meaning are presented, followed by Monte Carlo integration and random sampling, and by a general description of the main aspects and components of a typical Monte Carlo particle transport code. In particular, the most common biasing techniques are described, as well as the concepts of estimator and detector. After a discussion of the different types of errors, the issue of Quality Assurance is briefly considered.

  8. Calculation of extrapolation curves in the 4π(LS)β-γ coincidence technique with the Monte Carlo code Geant4.

    Science.gov (United States)

    Bobin, C; Thiam, C; Bouchard, J

    2016-03-01

    At LNE-LNHB, a liquid scintillation (LS) detection setup designed for Triple to Double Coincidence Ratio (TDCR) measurements is also used in the β-channel of a 4π(LS)β-γ coincidence system. This LS counter based on 3 photomultipliers was first modeled using the Monte Carlo code Geant4 to enable the simulation of optical photons produced by scintillation and Cerenkov effects. This stochastic modeling was especially designed for the calculation of double and triple coincidences between photomultipliers in TDCR measurements. In the present paper, this TDCR-Geant4 model is extended to 4π(LS)β-γ coincidence counting to enable the simulation of the efficiency-extrapolation technique by the addition of a γ-channel. This simulation tool aims at the prediction of systematic biases in activity determination due to eventual non-linearity of efficiency-extrapolation curves. First results are described in the case of the standardization (59)Fe. The variation of the γ-efficiency in the β-channel due to the Cerenkov emission is investigated in the case of the activity measurements of (54)Mn. The problem of the non-linearity between β-efficiencies is featured in the case of the efficiency tracing technique for the activity measurements of (14)C using (60)Co as a tracer.

  9. Monte Carlo calculation for the development of a BNCT neutron source (1eV-10KeV) using MCNP code.

    Science.gov (United States)

    El Moussaoui, F; El Bardouni, T; Azahra, M; Kamili, A; Boukhal, H

    2008-09-01

    Different materials have been studied in order to produce the epithermal neutron beam between 1eV and 10KeV, which are extensively used to irradiate patients with brain tumors such as GBM. For this purpose, we have studied three different neutrons moderators (H(2)O, D(2)O and BeO) and their combinations, four reflectors (Al(2)O(3), C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation showed that the best obtained assembly configuration corresponds to the combination of the three moderators H(2)O, BeO and D(2)O jointly to Al(2)O(3) reflector and two filter Cd+Bi optimize the spectrum of the epithermal neutron at 72%, and minimize the thermal neutron to 4% and thus it can be used to treat the deep tumor brain. The calculations have been performed by means of the Monte Carlo N (particle code MCNP 5C). Our results strongly encourage further studying of irradiation of the head with epithermal neutron fields.

  10. Pre- and post-processor for the wool won transport code

    CERN Document Server

    Fawley, W M

    2001-01-01

    ICOOL is a Fortran77 macroparticle transport code widely used by researchers to study the front end of a neutrino factory/muon collider. In part due to the desire that ICOOL be usable over multiple computer platforms and operating systems, the code uses simple text files for input/output services. This choice together with user-driven requests for greater and greater choice of lattice element type and configuration has led to ICOOL input decks becoming rather difficult to compose and modify easily. Moreover, the lack of a standard graphical postprocessor has prevented many ICOOL users from extracting all but the most simple results from the output files. Here I present two attempts to improve this situation: First, a simple but quite general graphical pre-processor (NIME) written in the Tcl/TK to permit users to write and maintain ASCII-formatted input files by use of simple macro definitions and expansions. Second, an interactive postprocessor written in Fortran90 and NCAR graphics, which allows users to def...

  11. Edge Transport Modeling using the 3D EMC3-Eirene code on Tokamaks and Stellarators

    Science.gov (United States)

    Lore, J. D.; Ahn, J. W.; Briesemeister, A.; Ferraro, N.; Labombard, B.; McLean, A.; Reinke, M.; Shafer, M.; Terry, J.

    2015-11-01

    The fluid plasma edge transport code EMC3-Eirene has been applied to aid data interpretation and understanding the results of experiments with 3D effects on several tokamaks. These include applied and intrinsic 3D magnetic fields, 3D plasma facing components, and toroidally and poloidally localized heat and particle sources. On Alcator C-Mod, a series of experiments explored the impact of toroidally and poloidally localized impurity gas injection on core confinement and asymmetries in the divertor fluxes, with the differences between the asymmetry in L-mode and H-mode qualitatively reproduced in the simulations due to changes in the impurity ionization in the private flux region. Modeling of NSTX experiments on the effect of 3D fields on detachment matched the trend of a higher density at which the detachment occurs when 3D fields are applied. On DIII-D, different magnetic field models were used in the simulation and compared against the 2D Thomson scattering diagnostic. In simulating each device different aspects of the code model are tested pointing to areas where the model must be further developed. The application to stellarator experiments will also be discussed. Work supported by U.S. DOE: DE-AC05-00OR22725, DE AC02-09CH11466, DE-FC02-99ER54512, and DE-FC02-04ER54698.

  12. EBQ code: transport of space-charge beams in axially symmetric devices

    Energy Technology Data Exchange (ETDEWEB)

    Paul, A.C.

    1982-11-01

    Such general-purpose space charge codes as EGUN, BATES, WOLF, and TRANSPORT do not gracefully accommodate the simulation of relativistic space-charged beams propagating a long distance in axially symmetric devices where a high degree of cancellation has occurred between the self-magnetic and self-electric forces of the beam. The EBQ code was written specifically to follow high current beam particles where space charge is important in long distance flight in axially symmetric machines possessing external electric and magnetic field. EBQ simultaneously tracks all trajectories so as to allow procedures for charge deposition based on inter-ray separations. The orbits are treated in Cartesian geometry (position and momentum) with z as the independent variable. Poisson's equation is solved in cylindrical geometry on an orthogonal rectangular mesh. EBQ can also handle problems involving multiple ion species where the space charge from each must be included. Such problems arise in the design of ion sources where different charge and mass states are present.

  13. Comparison of two numerical modelling codes for hydraulic and transport calculations in the near-field

    Energy Technology Data Exchange (ETDEWEB)

    Kalin, J., E-mail: jan.kalin@zag.s [Slovenian National Building and Civil Engineering Institute, Dimiceva 12, SI-1000 Ljubljana (Slovenia); Petkovsek, B., E-mail: borut.petkovsek@zag.s [Slovenian National Building and Civil Engineering Institute, Dimiceva 12, SI-1000 Ljubljana (Slovenia); Montarnal, Ph., E-mail: philippe.montarnal@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Genty, A., E-mail: alain.genty@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Deville, E., E-mail: estelle.deville@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Krivic, J., E-mail: jure.krivic@geo-zs.s [Geological Survey of Slovenia, Dimiceva 14, SI-1000 Ljubljana (Slovenia); Ratej, J., E-mail: joze.ratej@geo-zs.s [Geological Survey of Slovenia, Dimiceva 14, SI-1000 Ljubljana (Slovenia)

    2011-04-15

    In the past years the Slovenian Performance Analysis/Safety Assessment team has performed many generic studies for the future Slovenian low and intermediate level waste repository, most recently a Special Safety Analysis for the Krsko site. The modelling approach was to split the problem into three parts: near-field (detailed model of the repository), far-field (i.e., geosphere) and biosphere. In the Special Safety Analysis the code used to perform the near-field calculations was Hydrus2D. Recently the team has begun a cooperation with the French Commisariat al'Energie Atomique/Saclay (CEA/Saclay) and, as a part of this cooperation, began investigations into using the Alliances numerical platform for near-field calculations in order to compare the overall approach and calculated results. The article presents the comparison between these two codes for a silo-type repository that was considered in the Special Safety Analysis. The physical layout and characteristics of the repository are presented and a hydraulic and transport model of the repository is developed and implemented in Alliances. Some analysis of sensitivity to mesh fineness and to simulation timestep has been preformed and is also presented. The compared quantity is the output flux of radionuclides on the boundary of the model. Finally the results from Hydrus2D and Alliances are compared and the differences and similarities are commented.

  14. Applicability of the three-dimensional transport code Tort to the shielding analysis of the prototype FBR Monju

    Energy Technology Data Exchange (ETDEWEB)

    Takako, Shiraki [Mitsubishi Heavy Industries, Ltd (Japan); Shin, Usami; Zenro, Suzuoki; Takehide, Deshimaru [Japan Nuclear Cycle Development Institute (Japan); Kenji, Sasaki; Keiko, Tada; Hitoshi, Yokobori [Advanced Reactor Technology Co., Ltd (Japan)

    2003-07-01

    Shielding design of Monju was performed in 1980's by using the two-dimensional discrete ordinates transport code, DOT3.5. In view of complexity of the three-dimensional shielding geometry of Monju, the three-dimensional discrete ordinates transport code, TORT(2), has been applied to shielding measurement analyses of Monju in an attempt to prove practical usefulness of the code and to learn how much margin is associated with the shielding design performed by DOT3.5. This study has confirmed that TORT can practically be applied to the shielding measurement analyses of Monju, and has provided significant improvement in calculation accuracy thanks to its three-dimensional geometry employed, making the code applicable to the Monju shielding design evaluation analyses together with pre- and post-analyses of the shielding measurement now being planned. (authors)

  15. Self-consistent simulation of plasma scenarios for ITER using a combination of 1.5D transport codes and free-boundary equilibrium codes

    CERN Document Server

    Parail, V; Ambrosino, R; Artaud, J-F; Besseghir, K; Cavinato, M; Corrigan, G; Garcia, J; Garzotti, L; Gribov, Y; Imbeaux, F; Koechl, F; Labate, C V; Lister, J; Litaudon, X; Loarte, A; Maget, P; Mattei, M; McDonald, D; Nardon, E; Saibene, G; Sartori, R; Urban, J

    2013-01-01

    Self-consistent transport simulation of ITER scenarios is a very important tool for the exploration of the operational space and for scenario optimisation. It also provides an assessment of the compatibility of developed scenarios (which include fast transient events) with machine constraints, in particular with the poloidal field (PF) coil system, heating and current drive (H&CD), fuelling and particle and energy exhaust systems. This paper discusses results of predictive modelling of all reference ITER scenarios and variants using two suite of linked transport and equilibrium codes. The first suite consisting of the 1.5D core/2D SOL code JINTRAC [1] and the free boundary equilibrium evolution code CREATE-NL [2,3], was mainly used to simulate the inductive D-T reference Scenario-2 with fusion gain Q=10 and its variants in H, D and He (including ITER scenarios with reduced current and toroidal field). The second suite of codes was used mainly for the modelling of hybrid and steady state ITER scenarios. It...

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

    Energy Technology Data Exchange (ETDEWEB)

    Kostin, Mikhail [FRIB, MSU; Mokhov, Nikolai [FNAL; Niita, Koji [RIST, Japan

    2013-09-25

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

  17. Generic reactive transport codes as flexible tools to integrate soil organic matter degradation models with water, transport and geochemistry in soils

    Science.gov (United States)

    Jacques, Diederik; Gérard, Fréderic; Mayer, Uli; Simunek, Jirka; Leterme, Bertrand

    2016-04-01

    A large number of organic matter degradation, CO2 transport and dissolved organic matter models have been developed during the last decades. However, organic matter degradation models are in many cases strictly hard-coded in terms of organic pools, degradation kinetics and dependency on environmental variables. The scientific input of the model user is typically limited to the adjustment of input parameters. In addition, the coupling with geochemical soil processes including aqueous speciation, pH-dependent sorption and colloid-facilitated transport are not incorporated in many of these models, strongly limiting the scope of their application. Furthermore, the most comprehensive organic matter degradation models are combined with simplified representations of flow and transport processes in the soil system. We illustrate the capability of generic reactive transport codes to overcome these shortcomings. The formulations of reactive transport codes include a physics-based continuum representation of flow and transport processes, while biogeochemical reactions can be described as equilibrium processes constrained by thermodynamic principles and/or kinetic reaction networks. The flexibility of these type of codes allows for straight-forward extension of reaction networks, permits the inclusion of new model components (e.g.: organic matter pools, rate equations, parameter dependency on environmental conditions) and in such a way facilitates an application-tailored implementation of organic matter degradation models and related processes. A numerical benchmark involving two reactive transport codes (HPx and MIN3P) demonstrates how the process-based simulation of transient variably saturated water flow (Richards equation), solute transport (advection-dispersion equation), heat transfer and diffusion in the gas phase can be combined with a flexible implementation of a soil organic matter degradation model. The benchmark includes the production of leachable organic matter

  18. MCDB Monte Carlo dosimetry code system and its applications%MCDB蒙特卡罗剂量计算系统及应用

    Institute of Scientific and Technical Information of China (English)

    邓力; 李刚; 陈朝斌; 叶涛

    2012-01-01

    硼中子俘获治疗(BNCT)蒙特卡罗剂量计算软件系统MCDB(Monte Carlo dosimetry code for brain)已经开发成功.它包括医学前处理、剂量计算和后处理.前处理把CT、MRI图像数据自动转化为剂量计算的输入文件,剂量计算基于蒙特卡罗(MC)方法,后处理是确定照射方向和照射时间.为了提高剂量计算的精度和缩短计算时间,MCDB发展了针对体素模型的快速粒子径迹算法,构造材料矩阵和计数矩阵,程序实现了MPI并行化.通过一个病例,MCDB完成了从CT、MRI提取数据、剂量计算和后处理的全过程.计算取得了与MCNP程序一致的结果,串行计算速度较MCNP提高3倍以上,并行效率可以达到90%,完全满足临床对计算精度和计算时间的要求.%MCDB is developed for boron neutron capture therapy ( BNCT). This system consists of a medical pre-processor, a dose computation and a post-processor. MCDB automatically produces the input file from CT and MRI image data. In Monte Carlo dose calculation, several accelerated measures, such as the fast track technique , mesh tally matrix and material matrix, are developed. In this paper, we proposed a real model simulated by MCNP and MCDB, respectively. The almost same results as MCNP are achieved. MCDB is faster in computational speed than MCNP.

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

    Energy Technology Data Exchange (ETDEWEB)

    Randolph Schwarz; Leland L. Carter; Alysia Schwarz

    2005-08-23

    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.

  20. A Monte Carlo algorithm for degenerate plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Turrell, A.E., E-mail: a.turrell09@imperial.ac.uk; Sherlock, M.; Rose, S.J.

    2013-09-15

    A procedure for performing Monte Carlo calculations of plasmas with an arbitrary level of degeneracy is outlined. It has possible applications in inertial confinement fusion and astrophysics. Degenerate particles are initialised according to the Fermi–Dirac distribution function, and scattering is via a Pauli blocked binary collision approximation. The algorithm is tested against degenerate electron–ion equilibration, and the degenerate resistivity transport coefficient from unmagnetised first order transport theory. The code is applied to the cold fuel shell and alpha particle equilibration problem of inertial confinement fusion.

  1. 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.

  2. Analysis of the track- and dose-averaged LET and LET spectra in proton therapy using the GEANT4 Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Guan, Fada; Peeler, Christopher; Taleei, Reza; Randeniya, Sharmalee; Ge, Shuaiping; Mirkovic, Dragan; Mohan, Radhe; Titt, Uwe, E-mail: UTitt@mdanderson.org [Department of Radiation Physics, The University of Texas MD Anderson Cancer Center, 1515 Holcombe Boulevard, Houston, Texas 77030 (United States); Bronk, Lawrence [Department of Experimental Radiation Oncology, The University of Texas MD Anderson Cancer Center, 1515 Holcombe Boulevard, Houston, Texas 77030 (United States); Geng, Changran [Department of Nuclear Science and Engineering, Nanjing University of Aeronautics and Astronautics, Nanjing 210016, China and Department of Radiation Oncology, Massachusetts General Hospital and Harvard Medical School, Boston, Massachusetts 02114 (United States); Grosshans, David [Department of Experimental Radiation Oncology, The University of Texas MD Anderson Cancer Center, 1515 Holcombe Boulevard, Houston, Texas 77030 and Department of Radiation Oncology, The University of Texas MD Anderson Cancer Center, 1515 Holcombe Boulevard, Houston, Texas 77030 (United States)

    2015-11-15

    Purpose: The motivation of this study was to find and eliminate the cause of errors in dose-averaged linear energy transfer (LET) calculations from therapeutic protons in small targets, such as biological cell layers, calculated using the GEANT 4 Monte Carlo code. Furthermore, the purpose was also to provide a recommendation to select an appropriate LET quantity from GEANT 4 simulations to correlate with biological effectiveness of therapeutic protons. Methods: The authors developed a particle tracking step based strategy to calculate the average LET quantities (track-averaged LET, LET{sub t} and dose-averaged LET, LET{sub d}) using GEANT 4 for different tracking step size limits. A step size limit refers to the maximally allowable tracking step length. The authors investigated how the tracking step size limit influenced the calculated LET{sub t} and LET{sub d} of protons with six different step limits ranging from 1 to 500 μm in a water phantom irradiated by a 79.7-MeV clinical proton beam. In addition, the authors analyzed the detailed stochastic energy deposition information including fluence spectra and dose spectra of the energy-deposition-per-step of protons. As a reference, the authors also calculated the averaged LET and analyzed the LET spectra combining the Monte Carlo method and the deterministic method. Relative biological effectiveness (RBE) calculations were performed to illustrate the impact of different LET calculation methods on the RBE-weighted dose. Results: Simulation results showed that the step limit effect was small for LET{sub t} but significant for LET{sub d}. This resulted from differences in the energy-deposition-per-step between the fluence spectra and dose spectra at different depths in the phantom. Using the Monte Carlo particle tracking method in GEANT 4 can result in incorrect LET{sub d} calculation results in the dose plateau region for small step limits. The erroneous LET{sub d} results can be attributed to the algorithm to

  3. Evaluation of the response of a neutron detector of the Long-Counter type using a Monte Carlo transport simulation

    Energy Technology Data Exchange (ETDEWEB)

    Pazianotto, Mauricio Tizziani; Goncalez, Odair Lelis; Federico, Claudio Antonio [Centro Tecnico Aeroespacial (IEAv/CTA), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados; Carlson, Brett Vern [Centro Tecnico Aeroespacial (ITA/CTA), Sao Jose dos Campos, SP (Brazil). Inst. Tecnologico de Aeronautica

    2010-07-01

    Full text: The Institute for Advanced Studies (IEAv) is developing activities to study the dose levels of ionizing radiation from cosmic rays (CR) received by aircraft crews, sensitive equipment (on-board computers, for example) and embedded electronics in Brazilian airspace. Neutrons generated by the interaction of CR with the atmosphere are the dominant particles in the dose accumulation in electronic circuits and aircraft crews at flight altitude. Their production has a very broad energy spectrum, ranging from thermal neutrons (0.025eV ) to neutrons of several hundreds of MeV , making their detection a very difficult process. To observe the temporal variation in flow during the measurements, a detector of the Long Counter (LC) type is being used. This detector is designed to measure the one-way flow of neutrons with constant response over a wide energy range (thermal to 20 MeV ). However, to measure cosmic rays, the flow of which is non-directional, the dependence of the response on the angle of incidence, as well as energy, should be properly investigated. The objective of this study is to assess the angular response of the neutron detector (Long Counter) using the code MCNP5 (Monte Carlo N-Particle) and to compare it with the experimental data previously obtained with a {sup 241}Am-Be source at a distance of 1.66 m from the geometric center of the detector, varying the angle of incidence from 00 to 3600 in intervals of 150. The simulation was performed by modeling in detail the structure and materials of the LC, as well as the experimental arrangement for irradiation. The results of the simulation present reasonable agreement with the experimental data. This agreement shows that the modeling of the geometry of the source-detector system is adequate. The next step is to develop a model of neutron detection for the higher energy present in cosmic radiation fields, for which the experimental calibration is not so easily achievable. (author)

  4. Recent Improvements in the SHIELD-HIT Code

    DEFF Research Database (Denmark)

    Hansen, David Christoffer; Lühr, Armin Christian; Herrmann, Rochus

    2012-01-01

    Purpose: The SHIELD-HIT Monte Carlo particle transport code has previously been used to study a wide range of problems for heavy-ion treatment and has been benchmarked extensively against other Monte Carlo codes and experimental data. Here, an improved version of SHIELD-HIT is developed...... of using accelerator control files as a basis for the primaries. Furthermore, the code has been parallelized and efficiency is improved. The physical description of inelastic ion collisions has been modified. Results: The simulation of an experimental depth-dose distribution including a ripple filter...

  5. Validation of a commercial TPS based on the VMC(++) Monte Carlo code for electron beams: commissioning and dosimetric comparison with EGSnrc in homogeneous and heterogeneous phantoms.

    Science.gov (United States)

    Ferretti, A; Martignano, A; Simonato, F; Paiusco, M

    2014-02-01

    The aim of the present work was the validation of the VMC(++) Monte Carlo (MC) engine implemented in the Oncentra Masterplan (OMTPS) and used to calculate the dose distribution produced by the electron beams (energy 5-12 MeV) generated by the linear accelerator (linac) Primus (Siemens), shaped by a digital variable applicator (DEVA). The BEAMnrc/DOSXYZnrc (EGSnrc package) MC model of the linac head was used as a benchmark. Commissioning results for both MC codes were evaluated by means of 1D Gamma Analysis (2%, 2 mm), calculated with a home-made Matlab (The MathWorks) program, comparing the calculations with the measured profiles. The results of the commissioning of OMTPS were good [average gamma index (γ) > 97%]; some mismatches were found with large beams (size ≥ 15 cm). The optimization of the BEAMnrc model required to increase the beam exit window to match the calculated and measured profiles (final average γ > 98%). Then OMTPS dose distribution maps were compared with DOSXYZnrc with a 2D Gamma Analysis (3%, 3 mm), in 3 virtual water phantoms: (a) with an air step, (b) with an air insert, and (c) with a bone insert. The OMTPD and EGSnrc dose distributions with the air-water step phantom were in very high agreement (γ ∼ 99%), while for heterogeneous phantoms there were differences of about 9% in the air insert and of about 10-15% in the bone region. This is due to the Masterplan implementation of VMC(++) which reports the dose as "dose to water", instead of "dose to medium".

  6. Keno-Nr a Monte Carlo Code Simulating the Californium -252-SOURCE-DRIVEN Noise Analysis Experimental Method for Determining Subcriticality

    Science.gov (United States)

    Ficaro, Edward Patrick

    The ^{252}Cf -source-driven noise analysis (CSDNA) requires the measurement of the cross power spectral density (CPSD) G_ {23}(omega), between a pair of neutron detectors (subscripts 2 and 3) located in or near the fissile assembly, and the CPSDs, G_{12}( omega) and G_{13}( omega), between the neutron detectors and an ionization chamber 1 containing ^{252}Cf also located in or near the fissile assembly. The key advantage of this method is that the subcriticality of the assembly can be obtained from the ratio of spectral densities,{G _sp{12}{*}(omega)G_ {13}(omega)over G_{11 }(omega)G_{23}(omega) },using a point kinetic model formulation which is independent of the detector's properties and a reference measurement. The multigroup, Monte Carlo code, KENO-NR, was developed to eliminate the dependence of the measurement on the point kinetic formulation. This code utilizes time dependent, analog neutron tracking to simulate the experimental method, in addition to the underlying nuclear physics, as closely as possible. From a direct comparison of simulated and measured data, the calculational model and cross sections are validated for the calculation, and KENO-NR can then be rerun to provide a distributed source k_ {eff} calculation. Depending on the fissile assembly, a few hours to a couple of days of computation time are needed for a typical simulation executed on a desktop workstation. In this work, KENO-NR demonstrated the ability to accurately estimate the measured ratio of spectral densities from experiments using capture detectors performed on uranium metal cylinders, a cylindrical tank filled with aqueous uranyl nitrate, and arrays of safe storage bottles filled with uranyl nitrate. Good agreement was also seen between simulated and measured values of the prompt neutron decay constant from the fitted CPSDs. Poor agreement was seen between simulated and measured results using composite ^6Li-glass-plastic scintillators at large subcriticalities for the tank of

  7. Comparison of depth-dose distributions of proton therapeutic beams calculated by means of logical detectors and ionization chamber modeled in Monte Carlo codes

    Energy Technology Data Exchange (ETDEWEB)

    Pietrzak, Robert [Department of Nuclear Physics and Its Applications, Institute of Physics, University of Silesia, Katowice (Poland); Konefał, Adam, E-mail: adam.konefal@us.edu.pl [Department of Nuclear Physics and Its Applications, Institute of Physics, University of Silesia, Katowice (Poland); Sokół, Maria; Orlef, Andrzej [Department of Medical Physics, Maria Sklodowska-Curie Memorial Cancer Center, Institute of Oncology, Gliwice (Poland)

    2016-08-01

    The success of proton therapy depends strongly on the precision of treatment planning. Dose distribution in biological tissue may be obtained from Monte Carlo simulations using various scientific codes making it possible to perform very accurate calculations. However, there are many factors affecting the accuracy of modeling. One of them is a structure of objects called bins registering a dose. In this work the influence of bin structure on the dose distributions was examined. The MCNPX code calculations of Bragg curve for the 60 MeV proton beam were done in two ways: using simple logical detectors being the volumes determined in water, and using a precise model of ionization chamber used in clinical dosimetry. The results of the simulations were verified experimentally in the water phantom with Marcus ionization chamber. The average local dose difference between the measured relative doses in the water phantom and those calculated by means of the logical detectors was 1.4% at first 25 mm, whereas in the full depth range this difference was 1.6% for the maximum uncertainty in the calculations less than 2.4% and for the maximum measuring error of 1%. In case of the relative doses calculated with the use of the ionization chamber model this average difference was somewhat greater, being 2.3% at depths up to 25 mm and 2.4% in the full range of depths for the maximum uncertainty in the calculations of 3%. In the dose calculations the ionization chamber model does not offer any additional advantages over the logical detectors. The results provided by both models are similar and in good agreement with the measurements, however, the logical detector approach is a more time-effective method. - Highlights: • Influence of the bin structure on the proton dose distributions was examined for the MC simulations. • The considered relative proton dose distributions in water correspond to the clinical application. • MC simulations performed with the logical detectors and the

  8. Functional Coding Variation in Recombinant Inbred Mouse Lines Reveals Novel Serotonin Transporter-Associated Phenotypes

    Energy Technology Data Exchange (ETDEWEB)

    Carneiro, Ana [Vanderbilt University; Airey, David [University of Tennessee Health Science Center, Memphis; Thompson, Brent [Vanderbilt University; Zhu, C [Vanderbilt University; Rinchik, Eugene M [ORNL; Lu, Lu [University of Tennessee Health Science Center, Memphis; Chesler, Elissa J [ORNL; Erikson, Keith [University of North Carolina; Blakely, Randy [Vanderbilt University

    2009-01-01

    The human serotonin (5-hydroxytryptamine, 5-HT) transporter (hSERT, SLC6A4) figures prominently in the etiology or treatment of many prevalent neurobehavioral disorders including anxiety, alcoholism, depression, autism and obsessive-compulsive disorder (OCD). Here we utilize naturally occurring polymorphisms in recombinant inbred (RI) lines to identify novel phenotypes associated with altered SERT function. The widely used mouse strain C57BL/6J, harbors a SERT haplotype defined by two nonsynonymous coding variants (Gly39 and Lys152 (GK)). At these positions, many other mouse lines, including DBA/2J, encode Glu39 and Arg152 (ER haplotype), assignments found also in hSERT. Synaptosomal 5-HT transport studies revealed reduced uptake associated with the GK variant. Heterologous expression studies confirmed a reduced SERT turnover rate for the GK variant. Experimental and in silico approaches using RI lines (C57Bl/6J X DBA/2J=BXD) identifies multiple anatomical, biochemical and behavioral phenotypes specifically impacted by GK/ER variation. Among our findings are multiple traits associated with anxiety and alcohol consumption, as well as of the control of dopamine (DA) signaling. Further bioinformatic analysis of BXD phenotypes, combined with biochemical evaluation of SERT knockout mice, nominates SERT-dependent 5-HT signaling as a major determinant of midbrain iron homeostasis that, in turn, dictates ironregulated DA phenotypes. Our studies provide a novel example of the power of coordinated in vitro, in vivo and in silico approaches using murine RI lines to elucidate and quantify the system-level impact of gene variation.

  9. SU-E-CAMPUS-I-02: Estimation of the Dosimetric Error Caused by the Voxelization of Hybrid Computational Phantoms Using Triangle Mesh-Based Monte Carlo Transport

    Energy Technology Data Exchange (ETDEWEB)

    Lee, C [Division of Cancer Epidemiology and Genetics, National Cancer Institute, Bethesda, MD (United States); Badal, A [U.S. Food ' Drug Administration (CDRH/OSEL), Silver Spring, MD (United States)

    2014-06-15

    Purpose: Computational voxel phantom provides realistic anatomy but the voxel structure may result in dosimetric error compared to real anatomy composed of perfect surface. We analyzed the dosimetric error caused from the voxel structure in hybrid computational phantoms by comparing the voxel-based doses at different resolutions with triangle mesh-based doses. Methods: We incorporated the existing adult male UF/NCI hybrid phantom in mesh format into a Monte Carlo transport code, penMesh that supports triangle meshes. We calculated energy deposition to selected organs of interest for parallel photon beams with three mono energies (0.1, 1, and 10 MeV) in antero-posterior geometry. We also calculated organ energy deposition using three voxel phantoms with different voxel resolutions (1, 5, and 10 mm) using MCNPX2.7. Results: Comparison of organ energy deposition between the two methods showed that agreement overall improved for higher voxel resolution, but for many organs the differences were small. Difference in the energy deposition for 1 MeV, for example, decreased from 11.5% to 1.7% in muscle but only from 0.6% to 0.3% in liver as voxel resolution increased from 10 mm to 1 mm. The differences were smaller at higher energies. The number of photon histories processed per second in voxels were 6.4×10{sup 4}, 3.3×10{sup 4}, and 1.3×10{sup 4}, for 10, 5, and 1 mm resolutions at 10 MeV, respectively, while meshes ran at 4.0×10{sup 4} histories/sec. Conclusion: The combination of hybrid mesh phantom and penMesh was proved to be accurate and of similar speed compared to the voxel phantom and MCNPX. The lowest voxel resolution caused a maximum dosimetric error of 12.6% at 0.1 MeV and 6.8% at 10 MeV but the error was insignificant in some organs. We will apply the tool to calculate dose to very thin layer tissues (e.g., radiosensitive layer in gastro intestines) which cannot be modeled by voxel phantoms.

  10. Monte Carlo transition probabilities

    OpenAIRE

    Lucy, L. B.

    2001-01-01

    Transition probabilities governing the interaction of energy packets and matter are derived that allow Monte Carlo NLTE transfer codes to be constructed without simplifying the treatment of line formation. These probabilities are such that the Monte Carlo calculation asymptotically recovers the local emissivity of a gas in statistical equilibrium. Numerical experiments with one-point statistical equilibrium problems for Fe II and Hydrogen confirm this asymptotic behaviour. In addition, the re...

  11. Pre-conditioned Backward Monte Carlo solutions to radiative transport in planetary atmospheres. Fundamentals: Sampling of propagation directions in polarising media

    CERN Document Server

    Muñoz, García; Mills,; P, F

    2014-01-01

    Context. The interpretation of polarised radiation emerging from a planetary atmosphere must rely on solutions to the vector Radiative Transport Equation (vRTE). Monte Carlo integration of the vRTE is a valuable approach for its flexible treatment of complex viewing and/or illumination geometries and because it can intuitively incorporate elaborate physics. Aims. We present a novel Pre-Conditioned Backward Monte Carlo (PBMC) algorithm for solving the vRTE and apply it to planetary atmospheres irradiated from above. As classical BMC methods, our PBMC algorithm 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. Methods. We show that the neglect of polarisation in the sampling of photon propagation directions in classical BMC algorithms leads to unstable and biased solutions for conservative, optically-thick, strongly-polarising media such as Rayleigh atmospheres. The numerical difficulty is avoid...

  12. An Overview of the Monte Carlo Application ToolKit (MCATK)

    Energy Technology Data Exchange (ETDEWEB)

    Trahan, Travis John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-01-07

    MCATK is a C++ component-based Monte Carlo neutron-gamma transport software library designed to build specialized applications and designed to provide new functionality in existing general-purpose Monte Carlo codes like MCNP; it was developed with Agile software engineering methodologies under the motivation to reduce costs. The characteristics of MCATK can be summarized as follows: MCATK physics – continuous energy neutron-gamma transport with multi-temperature treatment, static eigenvalue (k and α) algorithms, time-dependent algorithm, fission chain algorithms; MCATK geometry – mesh geometries, solid body geometries. MCATK provides verified, unit-tested Monte Carlo components, flexibility in Monte Carlo applications development, and numerous tools such as geometry and cross section plotters. Recent work has involved deterministic and Monte Carlo analysis of stochastic systems. Static and dynamic analysis is discussed, and the results of a dynamic test problem are given.

  13. Application of the Finite Orbit Width Version of the CQL3D Code to Transport of Fast Ions

    Science.gov (United States)

    Petrov, Yu. V.; Harvey, R. W.

    2016-10-01

    The CQL3D bounce-averaged Fokker-Planck (FP) code now includes the ``fully'' neoclassical version in which the diffusion and advection processes are averaged over actual drift orbits, rather than using a 1st-order expansion. Incorporation of Finite-Orbit-Width (FOW) effects results in neoclassical radial transport caused by collisions, RF wave heating and by toroidal electric field (radial pinch). We apply the CQL3D-full-FOW code to study the thermalization and radial transport of high-energy particles, such as alpha-particles produced by fusion in ITER or deuterons from NBI in NSTX, under effect of their interaction with auxiliary RF waves. A particular attention is given to visualization of transport in 3D space of velocity +major-radius coordinates. Supported by USDOE Grants FC02-01ER54649, FG02-04ER54744, and SC0006614.

  14. Development and validation of MCNPX-based Monte Carlo treatment plan verification system

    OpenAIRE

    Iraj Jabbari; Shahram Monadi

    2015-01-01

    A Monte Carlo treatment plan verification (MCTPV) system was developed for clinical treatment plan verification (TPV), especially for the conformal and intensity-modulated radiotherapy (IMRT) plans. In the MCTPV, the MCNPX code was used for particle transport through the accelerator head and the patient body. MCTPV has an interface with TiGRT planning system and reads the information which is needed for Monte Carlo calculation transferred in digital image communications in medicine-radiation ...

  15. Estimating statistical uncertainty of Monte Carlo efficiency-gain in the context of a correlated sampling Monte Carlo code for brachytherapy treatment planning with non-normal dose distribution.

    Science.gov (United States)

    Mukhopadhyay, Nitai D; Sampson, Andrew J; Deniz, Daniel; Alm Carlsson, Gudrun; Williamson, Jeffrey; Malusek, Alexandr

    2012-01-01

    Correlated sampling Monte Carlo methods can shorten computing times in brachytherapy treatment planning. Monte Carlo efficiency is typically estimated via efficiency gain, defined as the reduction in computing time by correlated sampling relative to conventional Monte Carlo methods when equal statistical uncertainties have been achieved. The determination of the efficiency gain uncertainty arising from random effects, however, is not a straightforward task specially when the error distribution is non-normal. The purpose of this study is to evaluate the applicability of the F distribution and standardized uncertainty propagation methods (widely used in metrology to estimate uncertainty of physical measurements) for predicting confidence intervals about efficiency gain estimates derived from single Monte Carlo runs using fixed-collision correlated sampling in a simplified brachytherapy geometry. A bootstrap based algorithm was used to simulate the probability distribution of the efficiency gain estimates and the shortest 95% confidence interval was estimated from this distribution. It was found that the corresponding relative uncertainty was as large as 37% for this particular problem. The uncertainty propagation framework predicted confidence intervals reasonably well; however its main disadvantage was that uncertainties of input quantities had to be calculated in a separate run via a Monte Carlo method. The F distribution noticeably underestimated the confidence interval. These discrepancies were influenced by several photons with large statistical weights which made extremely large contributions to the scored absorbed dose difference. The mechanism of acquiring high statistical weights in the fixed-collision correlated sampling method was explained and a mitigation strategy was proposed.

  16. Estimating statistical uncertainty of Monte Carlo efficiency-gain in the context of a correlated sampling Monte Carlo code for brachytherapy treatment planning with non-normal dose distribution

    Energy Technology Data Exchange (ETDEWEB)

    Mukhopadhyay, Nitai D. [Department of Biostatistics, Virginia Commonwealth University, Richmond, VA 23298 (United States); Sampson, Andrew J. [Department of Radiation Oncology, Virginia Commonwealth University, Richmond, VA 23298 (United States); Deniz, Daniel; Alm Carlsson, Gudrun [Department of Radiation Physics, Faculty of Health Sciences, Linkoeping University, SE 581 85 (Sweden); Williamson, Jeffrey [Department of Radiation Oncology, Virginia Commonwealth University, Richmond, VA 23298 (United States); Malusek, Alexandr, E-mail: malusek@ujf.cas.cz [Department of Radiation Physics, Faculty of Health Sciences, Linkoeping University, SE 581 85 (Sweden); Department of Radiation Dosimetry, Nuclear Physics Institute AS CR v.v.i., Na Truhlarce 39/64, 180 86 Prague (Czech Republic)

    2012-01-15

    Correlated sampling Monte Carlo methods can shorten computing times in brachytherapy treatment planning. Monte Carlo efficiency is typically estimated via efficiency gain, defined as the reduction in computing time by correlated sampling relative to conventional Monte Carlo methods when equal statistical uncertainties have been achieved. The determination of the efficiency gain uncertainty arising from random effects, however, is not a straightforward task specially when the error distribution is non-normal. The purpose of this study is to evaluate the applicability of the F distribution and standardized uncertainty propagation methods (widely used in metrology to estimate uncertainty of physical measurements) for predicting confidence intervals about efficiency gain estimates derived from single Monte Carlo runs using fixed-collision correlated sampling in a simplified brachytherapy geometry. A bootstrap based algorithm was used to simulate the probability distribution of the efficiency gain estimates and the shortest 95% confidence interval was estimated from this distribution. It was found that the corresponding relative uncertainty was as large as 37% for this particular problem. The uncertainty propagation framework predicted confidence intervals reasonably well; however its main disadvantage was that uncertainties of input quantities had to be calculated in a separate run via a Monte Carlo method. The F distribution noticeably underestimated the confidence interval. These discrepancies were influenced by several photons with large statistical weights which made extremely large contributions to the scored absorbed dose difference. The mechanism of acquiring high statistical weights in the fixed-collision correlated sampling method was explained and a mitigation strategy was proposed.

  17. MCNP{trademark} Monte Carlo: A precis of MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Adams, K.J.

    1996-06-01

    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.

  18. Coupling an analytical description of anti-scatter grids with simulation software of radiographic systems using Monte Carlo code; Couplage d'une methode de description analytique de grilles anti diffusantes avec un logiciel de simulation de systemes radiographiques base sur un code Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Rinkel, J.; Dinten, J.M.; Tabary, J

    2004-07-01

    The use of focused anti-scatter grids on digital radiographic systems with two-dimensional detectors produces acquisitions with a decreased scatter to primary ratio and thus improved contrast and resolution. Simulation software is of great interest in optimizing grid configuration according to a specific application. Classical simulators are based on complete detailed geometric descriptions of the grid. They are accurate but very time consuming since they use Monte Carlo code to simulate scatter within the high-frequency grids. We propose a new practical method which couples an analytical simulation of the grid interaction with a radiographic system simulation program. First, a two dimensional matrix of probability depending on the grid is created offline, in which the first dimension represents the angle of impact with respect to the normal to the grid lines and the other the energy of the photon. This matrix of probability is then used by the Monte Carlo simulation software in order to provide the final scattered flux image. To evaluate the gain of CPU time, we define the increasing factor as the increase of CPU time of the simulation with as opposed to without the grid. Increasing factors were calculated with the new model and with classical methods representing the grid with its CAD model as part of the object. With the new method, increasing factors are shorter by one to two orders of magnitude compared with the second one. These results were obtained with a difference in calculated scatter of less than five percent between the new and the classical method. (authors)

  19. Energy and resolution calibration of NaI(Tl) and LaBr{sub 3}(Ce) scintillators and validation of an EGS5 Monte Carlo user code for efficiency calculations

    Energy Technology Data Exchange (ETDEWEB)

    Casanovas, R., E-mail: ramon.casanovas@urv.cat [Unitat de Fisica Medica, Facultat de Medicina i Ciencies de la Salut, Universitat Rovira i Virgili, ES-43201 Reus (Tarragona) (Spain); Morant, J.J. [Servei de Proteccio Radiologica, Facultat de Medicina i Ciencies de la Salut, Universitat Rovira i Virgili, ES-43201 Reus (Tarragona) (Spain); Salvado, M. [Unitat de Fisica Medica, Facultat de Medicina i Ciencies de la Salut, Universitat Rovira i Virgili, ES-43201 Reus (Tarragona) (Spain)

    2012-05-21

    The radiation detectors yield the optimal performance if they are accurately calibrated. This paper presents the energy, resolution and efficiency calibrations for two scintillation detectors, NaI(Tl) and LaBr{sub 3}(Ce). For the two former calibrations, several fitting functions were tested. To perform the efficiency calculations, a Monte Carlo user code for the EGS5 code system was developed with several important implementations. The correct performance of the simulations was validated by comparing the simulated spectra with the experimental spectra and reproducing a number of efficiency and activity calculations. - Highlights: Black-Right-Pointing-Pointer NaI(Tl) and LaBr{sub 3}(Ce) scintillation detectors are used for gamma-ray spectrometry. Black-Right-Pointing-Pointer Energy, resolution and efficiency calibrations are discussed for both detectors. Black-Right-Pointing-Pointer For the two former calibrations, several fitting functions are tested. Black-Right-Pointing-Pointer A Monte Carlo user code for EGS5 was developed for the efficiency calculations. Black-Right-Pointing-Pointer The code was validated reproducing some efficiency and activity calculations.

  20. Suite of Benchmark Tests to Conduct Mesh-Convergence Analysis of Nonlinear and Non-constant Coefficient Transport Codes

    Science.gov (United States)

    Zamani, K.; Bombardelli, F. A.

    2014-12-01

    Verification of geophysics codes is imperative to avoid serious academic as well as practical consequences. In case that access to any given source code is not possible, the Method of Manufactured Solution (MMS) cannot be employed in code verification. In contrast, employing the Method of Exact Solution (MES) has several practical advantages. In this research, we first provide four new one-dimensional analytical solutions designed for code verification; these solutions are able to uncover the particular imperfections of the Advection-diffusion-reaction equation, such as nonlinear advection, diffusion or source terms, as well as non-constant coefficient equations. After that, we provide a solution of Burgers' equation in a novel setup. Proposed solutions satisfy the continuity of mass for the ambient flow, which is a crucial factor for coupled hydrodynamics-transport solvers. Then, we use the derived analytical solutions for code verification. To clarify gray-literature issues in the verification of transport codes, we designed a comprehensive test suite to uncover any imperfection in transport solvers via a hierarchical increase in the level of tests' complexity. The test suite includes hundreds of unit tests and system tests to check vis-a-vis the portions of the code. Examples for checking the suite start by testing a simple case of unidirectional advection; then, bidirectional advection and tidal flow and build up to nonlinear cases. We design tests to check nonlinearity in velocity, dispersivity and reactions. The concealing effect of scales (Peclet and Damkohler numbers) on the mesh-convergence study and appropriate rem