Parallel processing Monte Carlo radiation transport codes
Issues related to distributed-memory multiprocessing as applied to Monte Carlo radiation transport are discussed. Measurements of communication overhead are presented for the radiation transport code MCNP which employs the communication software package PVM, and average efficiency curves are provided for a homogeneous virtual machine
The MCNPX Monte Carlo Radiation Transport Code
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 MCNP4c 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
THE MCNPX MONTE CARLO RADIATION TRANSPORT CODE
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.
Morse Monte Carlo Radiation Transport Code System
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)
MORSE Monte Carlo radiation transport code system
This report is an addendum to the MORSE report, ORNL-4972, originally published in 1975. This addendum contains descriptions of several modifications to the MORSE Monte Carlo Code, replacement pages containing corrections, Part II of the report which was previously unpublished, and a new Table of Contents. The modifications include a Klein Nishina estimator for gamma rays. Use of such an estimator required changing the cross section routines to process pair production and Compton scattering cross sections directly from ENDF tapes and writing a new version of subroutine RELCOL. Another modification is the use of free form input for the SAMBO analysis data. This required changing subroutines SCORIN and adding new subroutine RFRE. References are updated, and errors in the original report have been corrected
A New Monte Carlo Neutron Transport Code at UNIST
Monte Carlo neutron transport code named MCS is under development at UNIST for the advanced reactor design and research purpose. This MC code can be used for fixed source calculation and criticality calculation. Continuous energy neutron cross section data and multi-group cross section data can be used for the MC calculation. This paper presents the overview of developed MC code and its calculation results. The real time fixed source calculation ability is also tested in this paper. The calculation results show good agreement with commercial code and experiment. A new Monte Carlo neutron transport code is being developed at UNIST. The MC codes are tested with several benchmark problems: ICSBEP, VENUS-2, and Hoogenboom-Martin benchmark. These benchmarks covers pin geometry to 3-dimensional whole core, and results shows good agreement with reference results
Current status of the PSG Monte Carlo neutron transport code
PSG is a new Monte Carlo neutron transport code, developed at the Technical Research Centre of Finland (VTT). The code is mainly intended for fuel assembly-level reactor physics calculations, such as group constant generation for deterministic reactor simulator codes. This paper presents the current status of the project and the essential capabilities of the code. Although the main application of PSG is in lattice calculations, the geometry is not restricted in two dimensions. This paper presents the validation of PSG against the experimental results of the three-dimensional MOX fuelled VENUS-2 reactor dosimetry benchmark. (authors)
MORSE Monte Carlo radiation transport code system
For a number of years the MORSE user community has requested additional help in setting up problems using various options. The sample problems distributed with MORSE did not fully demonstrate the capability of the code. At Oak Ridge National Laboratory the code originators had a complete set of sample problems, but funds for documenting and distributing them were never available. Recently the number of requests for listings of input data and results for running some particular option the user was trying to implement has increased to the point where it is not feasible to handle them on an individual basis. Consequently it was decided to package a set of sample problems which illustrates more adequately how to run MORSE. This write-up may be added to Part III of the MORSE report. These sample problems include a combined neutron-gamma case, a neutron only case, a gamma only case, an adjoint case, a fission case, a time-dependent fission case, the collision density case, an XCHEKR run and a PICTUR run
TRIPOLI-3: a neutron/photon Monte Carlo transport code
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.)
Acceleration of a Monte Carlo radiation transport code
Execution time for the Integrated TIGER Series (ITS) Monte Carlo radiation transport code has been reduced by careful re-coding of computationally intensive subroutines. Three test cases for the TIGER (1-D slab geometry), CYLTRAN (2-D cylindrical geometry), and ACCEPT (3-D arbitrary geometry) codes were identified and used to benchmark and profile program execution. Based upon these results, sixteen top time-consuming subroutines were examined and nine of them modified to accelerate computations with equivalent numerical output to the original. The results obtained via this study indicate that speedup factors of 1.90 for the TIGER code, 1.67 for the CYLTRAN code, and 1.11 for the ACCEPT code are achievable. copyright 1996 American Institute of Physics
Verification of Monte Carlo transport codes by activation experiments
With the increasing energies and intensities of heavy-ion accelerator facilities, the problem of an excessive activation of the accelerator components caused by beam losses becomes more and more important. Numerical experiments using Monte Carlo transport codes are performed in order to assess the levels of activation. The heavy-ion versions of the codes were released approximately a decade ago, therefore the verification is needed to be sure that they give reasonable results. Present work is focused on obtaining the experimental data on activation of the targets by heavy-ion beams. Several experiments were performed at GSI Helmholtzzentrum fuer Schwerionenforschung. The interaction of nitrogen, argon and uranium beams with aluminum targets, as well as interaction of nitrogen and argon beams with copper targets was studied. After the irradiation of the targets by different ion beams from the SIS18 synchrotron at GSI, the γ-spectroscopy analysis was done: the γ-spectra of the residual activity were measured, the radioactive nuclides were identified, their amount and depth distribution were detected. The obtained experimental results were compared with the results of the Monte Carlo simulations using FLUKA, MARS and SHIELD. The discrepancies and agreements between experiment and simulations are pointed out. The origin of discrepancies is discussed. Obtained results allow for a better verification of the Monte Carlo transport codes, and also provide information for their further development. The necessity of the activation studies for accelerator applications is discussed. The limits of applicability of the heavy-ion beam-loss criteria were studied using the FLUKA code. FLUKA-simulations were done to determine the most preferable from the radiation protection point of view materials for use in accelerator components.
Parallelization of a Monte Carlo particle transport simulation code
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.
KAMCCO, a reactor physics Monte Carlo neutron transport code
KAMCCO is a 3-dimensional reactor Monte Carlo code for fast neutron physics problems. Two options are available for the solution of 1) the inhomogeneous time-dependent neutron transport equation (census time scheme), and 2) the homogeneous static neutron transport equation (generation cycle scheme). The user defines the desired output, e.g. estimates of reaction rates or neutron flux integrated over specified volumes in phase space and time intervals. Such primary quantities can be arbitrarily combined, also ratios of these quantities can be estimated with their errors. The Monte Carlo techniques are mostly analogue (exceptions: Importance sampling for collision processes, ELP/MELP, Russian roulette and splitting). Estimates are obtained from the collision and track length estimators. Elastic scattering takes into account first order anisotropy in the center of mass system. Inelastic scattering is processed via the evaporation model or via the excitation of discrete levels. For the calculation of cross sections, the energy is treated as a continuous variable. They are computed by a) linear interpolation, b) from optionally Doppler broadened single level Breit-Wigner resonances or c) from probability tables (in the region of statistically distributed resonances). (orig.)
Verification of Monte Carlo transport codes FLUKA, Mars and Shield
The present study is a continuation of the project 'Verification of Monte Carlo Transport Codes' which is running at GSI as a part of activation studies of FAIR relevant materials. It includes two parts: verification of stopping modules of FLUKA, MARS and SHIELD-A (with ATIMA stopping module) and verification of their isotope production modules. The first part is based on the measurements of energy deposition function of uranium ions in copper and stainless steel. The irradiation was done at 500 MeV/u and 950 MeV/u, the experiment was held at GSI from September 2004 until May 2005. The second part is based on gamma-activation studies of an aluminium target irradiated with an argon beam of 500 MeV/u in August 2009. Experimental depth profiling of the residual activity of the target is compared with the simulations. (authors)
Parallel implementation of the Monte Carlo transport code EGS4 on the hypercube
Monte Carlo transport codes are commonly used in the study of particle interactions. The CALOR89 code system is a combination of several Monte Carlo transport and analysis programs. In order to produce good results, a typical Monte Carlo run will have to produce many particle histories. On a single processor computer, the transport calculation can take a huge amount of time. However, if the transport of particles were divided among several processors in a multiprocessor machine, the time can be drastically reduced
Vectorization techniques for neutron transport Monte Carlo codes
Four Monte Carlo codes, KENO IV, MORSE-DD, MCNP and VIM, have been vectorized already at JAERI Computing Center aiming at an increase in clculation performance, and speed-up ratios of vectorized codes to the original ones were found to be low values between 1.3 and 1.5. In this report the vectorization processes for these four codes are reviewed comprehensively, and methods of analysis for vectorization, modification of control structures of codes and debugging techniques are discussed. The reason for low speed-up ratios is also discussed. (author)
A comparison between the Monte Carlo radiation transport codes MCNP and MCBEND
Sawamura, Hidenori; Nishimura, Kazuya [Computer Software Development Co., Ltd., Tokyo (Japan)
2001-01-01
In Japan, almost of all radiation analysts are using the MCNP code and MVP code on there studies. But these codes have not had automatic variance reduction. MCBEND code made by UKAEA have automatic variance reduction. And, MCBEND code is user friendly more than other Monte Carlo Radiation Transport Codes. Our company was first introduced MCBEND code in Japan. Therefore, we compared with MCBEND code and MCNP code about functions and production capacity. (author)
MCNP, a general Monte Carlo code for neutron and photon transport: a summary
The general-purpose Monte Carlo code MCNP can be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces
A vectorized Monte Carlo code for modeling photon transport in SPECT
A vectorized Monte Carlo computer code has been developed for modeling photon transport in single photon emission computed tomography (SPECT). The code models photon transport in a uniform attenuating region and photon detection by a gamma camera. It is adapted from a history-based Monte Carlo code in which photon history data are stored in scalar variables and photon histories are computed sequentially. The vectorized code is written in FORTRAN77 and uses an event-based algorithm in which photon history data are stored in arrays and photon history computations are performed within DO loops. The indices of the DO loops range over the number of photon histories, and these loops may take advantage of the vector processing unit of our Stellar GS1000 computer for pipelined computations. Without the use of the vector processor the event-based code is faster than the history-based code because of numerical optimization performed during conversion to the event-based algorithm. When only the detection of unscattered photons is modeled, the event-based code executes 5.1 times faster with the use of the vector processor than without; when the detection of scattered and unscattered photons is modeled the speed increase is a factor of 2.9. Vectorization is a valuable way to increase the performance of Monte Carlo code for modeling photon transport in SPECT
The general purpose code BALTORO was written for coupling the three-dimensional Monte-Carlo /MC/ with the one-dimensional Discrete Ordinates /DO/ radiation transport calculations. The quantity of a radiation-induced /neutrons or gamma-rays/ nuclear effect or the score from a radiation-yielding nuclear effect can be analysed in this way. (author)
ITS - The integrated TIGER series of coupled electron/photon Monte Carlo transport codes
The TIGER series of time-independent coupled electron/photon Monte Carlo transport codes is a group of multimaterial, multidimensional codes designed to provide a state-of-the-art description of the production and transport of the electron/photon cascade. The codes follow both electrons and photons from 1.0 GeV down to 1.0 keV, and the user has the option of combining the collisional transport with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence. Source particles can be either electrons or photons. The most important output data are (a) charge and energy deposition profiles, (b) integral and differential escape coefficients for both electrons and photons, (c) differential electron and photon flux, and (d) pulse-height distributions for selected regions of the problem geometry. The base codes of the series differ from one another primarily in their dimensionality and geometric modeling. They include (a) a one-dimensional multilayer code, (b) a code that describes the transport in two-dimensional axisymmetric cylindrical material geometries with a fully three-dimensional description of particle trajectories, and (c) a general three-dimensional transport code which employs a combinatorial geometry scheme. These base codes were designed primarily for describing radiation transport for those situations in which the detailed atomic structure of the transport medium is not important. For some applications, it is desirable to have a more detailed model of the low energy transport. The system includes three additional codes that contain a more elaborate ionization/relaxation model than the base codes. Finally, the system includes two codes that combine the collisional transport of the multidimensional base codes with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence
Parallelization of MCATNP MONTE CARLO particle transport code by using MPI
A Monte Carlo code for simulating Atmospheric Transport of Neutrons and Photons (MCATNP) is used to simulate the ionization effects caused by high altitude nuclear detonation (HAND) and it was parallelized in MPI by adopting the leap random number producer and modifying the original serial code. The parallel results and serial results are identical. The speedup increases almost linearly with the number of processors used. The parallel efficiency is up to to 97% while 16 processors are used, and 94% while 32 are used. The experimental results show that parallelization can obviously reduce the calculation time of Monte Carlo simulation of HAND ionization effects. (authors)
Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)
The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided
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...
COG is a major multiparticle simulation code in the LLNL Monte Carlo radiation transport toolkit. It was designed to solve deep-penetration radiation shielding problems in arbitrarily complex 3D geometries, involving coupled transport of photons, neutrons, and electrons. COG was written to provide as much accuracy as the underlying cross-sections will allow, and has a number of variance-reduction features to speed computations. Recently COG has been applied to the simulation of high- resolution radiographs of complex objects and the evaluation of contraband detection schemes. In this paper we will give a brief description of the capabilities of the COG transport code and show several examples of neutron and gamma-ray imaging simulations. Keywords: Monte Carlo, radiation transport, simulated radiography, nonintrusive inspection, neutron imaging
Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes
Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code)
Modelling photon transport in non-uniform media for SPECT with a vectorized Monte Carlo code.
Smith, M F
1993-10-01
A vectorized Monte Carlo code has been developed for modelling photon transport in non-uniform media for single-photon-emission computed tomography (SPECT). The code is designed to compute photon detection kernels, which are used to build system matrices for simulating SPECT projection data acquisition and for use in matrix-based image reconstruction. Non-uniform attenuating and scattering regions are constructed from simple three-dimensional geometric shapes, in which the density and mass attenuation coefficients are individually specified. On a Stellar GS1000 computer, Monte Carlo simulations are performed between 1.6 and 2.0 times faster when the vector processor is utilized than when computations are performed in scalar mode. Projection data acquired with a clinical SPECT gamma camera for a line source in a non-uniform thorax phantom are well modelled by Monte Carlo simulations. The vectorized Monte Carlo code was used to stimulate a 99Tcm SPECT myocardial perfusion study, and compensations for non-uniform attenuation and the detection of scattered photons improve activity estimation. The speed increase due to vectorization makes Monte Carlo simulation more attractive as a tool for modelling photon transport in non-uniform media for SPECT. PMID:8248288
Subroutines to Simulate Fission Neutrons for Monte Carlo Transport Codes
Lestone, J P
2014-01-01
Fortran subroutines have been written to simulate the production of fission neutrons from the spontaneous fission of 252Cf and 240Pu, and from the thermal neutron induced fission of 239Pu and 235U. The names of these four subroutines are getnv252, getnv240, getnv239, and getnv235, respectively. These subroutines reproduce measured first, second, and third moments of the neutron multiplicity distributions, measured neutron-fission correlation data for the spontaneous fission of 252Cf, and measured neutron-neutron correlation data for both the spontaneous fission of 252Cf and the thermal neutron induced fission of 235U. The codes presented here can be used to study the possible uses of neutron-neutron correlations in the area of transparency measurements and the uses of neutron-neutron correlations in coincidence neutron imaging.
Parallel processing of Monte Carlo code MCNP for particle transport problem
Higuchi, Kenji; Kawasaki, Takuji
1996-06-01
It is possible to vectorize or parallelize Monte Carlo codes (MC code) for photon and neutron transport problem, making use of independency of the calculation for each particle. Applicability of existing MC code to parallel processing is mentioned. As for parallel computer, we have used both vector-parallel processor and scalar-parallel processor in performance evaluation. We have made (i) vector-parallel processing of MCNP code on Monte Carlo machine Monte-4 with four vector processors, (ii) parallel processing on Paragon XP/S with 256 processors. In this report we describe the methodology and results for parallel processing on two types of parallel or distributed memory computers. In addition, we mention the evaluation of parallel programming environments for parallel computers used in the present work as a part of the work developing STA (Seamless Thinking Aid) Basic Software. (author)
PyMercury: Interactive Python for the Mercury Monte Carlo Particle Transport Code
Iandola, F N; O' Brien, M J; Procassini, R J
2010-11-29
Monte Carlo particle transport applications are often written in low-level languages (C/C++) for optimal performance on clusters and supercomputers. However, this development approach often sacrifices straightforward usability and testing in the interest of fast application performance. To improve usability, some high-performance computing applications employ mixed-language programming with high-level and low-level languages. In this study, we consider the benefits of incorporating an interactive Python interface into a Monte Carlo application. With PyMercury, a new Python extension to the Mercury general-purpose Monte Carlo particle transport code, we improve application usability without diminishing performance. In two case studies, we illustrate how PyMercury improves usability and simplifies testing and validation in a Monte Carlo application. In short, PyMercury demonstrates the value of interactive Python for Monte Carlo particle transport applications. In the future, we expect interactive Python to play an increasingly significant role in Monte Carlo usage and testing.
PyMercury: Interactive Python for the Mercury Monte Carlo Particle Transport Code
Monte Carlo particle transport applications are often written in low-level languages (C/C++) for optimal performance on clusters and supercomputers. However, this development approach often sacrifices straightforward usability and testing in the interest of fast application performance. To improve usability, some high-performance computing applications employ mixed-language programming with high-level and low-level languages. In this study, we consider the benefits of incorporating an interactive Python interface into a Monte Carlo application. With PyMercury, a new Python extension to the Mercury general-purpose Monte Carlo particle transport code, we improve application usability without diminishing performance. In two case studies, we illustrate how PyMercury improves usability and simplifies testing and validation in a Monte Carlo application. In short, PyMercury demonstrates the value of interactive Python for Monte Carlo particle transport applications. In the future, we expect interactive Python to play an increasingly significant role in Monte Carlo usage and testing.
MCNP: a general Monte Carlo code for neutron and photon transport
MCNP is a very general Monte Carlo neutron photon transport code system with approximately 250 person years of Group X-6 code development invested. It is extremely portable, user-oriented, and a true production code as it is used about 60 Cray hours per month by about 150 Los Alamos users. It has as its data base the best cross-section evaluations available. MCNP contains state-of-the-art traditional and adaptive Monte Carlo techniques to be applied to the solution of an ever-increasing number of problems. Excellent user-oriented documentation is available for all facets of the MCNP code system. Many useful and important variants of MCNP exist for special applications. The Radiation Shielding Information Center (RSIC) in Oak Ridge, Tennessee is the contact point for worldwide MCNP code and documentation distribution. A much improved MCNP Version 3A will be available in the fall of 1985, along with new and improved documentation. Future directions in MCNP development will change the meaning of MCNP to Monte Carlo N Particle where N particle varieties will be transported
Srna - Monte Carlo codes for proton transport simulation in combined and voxelized geometries
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.
Srna-Monte Carlo codes for proton transport simulation in combined and voxelized geometries
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. (author)
The use of Monte Carlo radiation transport codes in radiation physics and dosimetry
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...
Penelope - A code system for Monte Carlo simulation of electron and photon transport
The computer code system PENELOPE (version 2001) performs Monte Carlo simulation of coupled electron-photon transport in arbitrary materials for a wide energy range, from a few hundred eV to about 1 GeV. 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 geometry package called PENGEOM 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 PENELOPE code system, but also to provide the user with the necessary information to understand the details of the Monte-Carlo algorithm. (authors)
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.
MC++: A parallel, portable, Monte Carlo neutron transport code in C++
MC++ is an implicit multi-group Monte Carlo neutron transport code written in C++ and based on the Parallel Object-Oriented Methods and Applications (POOMA) class library. MC++ runs in parallel on and is portable to a wide variety of platforms, including MPPs, SMPs, and clusters of UNIX workstations. MC++ is being developed to provide transport capabilities to the Accelerated Strategic Computing Initiative (ASCI). It is also intended to form the basis of the first transport physics framework (TPF), which is a C++ class library containing appropriate abstractions, objects, and methods for the particle transport problem. The transport problem is briefly described, as well as the current status and algorithms in MC++ for solving the transport equation. The alpha version of the POOMA class library is also discussed, along with the implementation of the transport solution algorithms using POOMA. Finally, a simple test problem is defined and performance and physics results from this problem are discussed on a variety of platforms
A portable, parallel, object-oriented Monte Carlo neutron transport code in C++
We have developed a multi-group Monte Carlo neutron transport code using C++ and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and α-eigenvalues and is portable to and runs parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities of MC++ are discussed, along with physics and performance results on a variety of hardware, including all Accelerated Strategic Computing Initiative (ASCI) hardware. Current parallel performance indicates the ability to compute α-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed
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.
A new assembly-level Monte Carlo neutron transport code for reactor physics calculations
This paper presents a new assembly-level Monte Carlo neutron transport code, specifically intended for diffusion code group-constant generation and other reactor physics calculations. The code is being developed at the Technical Research Centre of Finland (VTT), under the working title 'Probabilistic Scattering Game', or PSG. The PSG code uses a method known as Woodcock tracking to simulate neutron histories. The advantages of the method include fast simulation in complex geometries and relatively simple handling of complicated geometrical objects. The main drawback is the inability to calculate reaction rates in optically thin volumes. This narrows the field of application to calculations involving parameters integrated over large volumes. The main features of the PSG code and the Woodcock tracking method are introduced. The code is applied in three example cases, involving infinite lattices of two-dimensional LWR fuel assemblies. Comparison calculations are carried out using MCNP4C and CASMO-4E. The results reveal that the code performs quite well in the calculation cases of this study, especially when compared to MCNP. The PSG code is still under extensive development and there are both flaws in the simulation of the interaction physics and programming errors in the source code. The results presented here, however, seem very encouraging, especially considering the early development stage of the code. (author)
This paper discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package developed and maintained at Oak Ridge National Laboratory. It has been developed to scale well from laptop to small computing clusters to advanced supercomputers. Special features of Shift include hybrid capabilities for variance reduction such as CADIS and FW-CADIS, and advanced parallel decomposition and tally methods optimized for scalability on supercomputing architectures. Shift has been validated and verified against various reactor physics benchmarks and compares well to other state-of-the-art Monte Carlo radiation transport codes such as MCNP5, CE KENO-VI, and OpenMC. Some specific benchmarks used for verification and validation include the CASL VERA criticality test suite and several Westinghouse AP1000® problems. These benchmark and scaling studies show promising results
Srna-Monte Carlo codes for proton transport simulation in combined and voxelized geometries
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...
ITS Version 6 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes.
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.
Overview of TRIPOLI-4 version 7, Continuous-energy Monte Carlo Transport Code
The TRIPOLI-4 code is used essentially for four major classes of applications: shielding studies, criticality studies, core physics studies, and instrumentation studies. In this updated overview of the Monte Carlo transport code TRIPOLI-4, we list and describe its current main features, including recent developments or extended capacities like effective beta estimation, photo-nuclear reactions or extended mesh tallies. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. TRIPOLI-4 has support for execution in parallel mode. Special features and applications are also presented concerning: 'particles storage', resuming a stopped TRIPOLI-4 run, collision bands, Green's functions, source convergence in criticality mode, and mesh tally
Application of ENDF nuclear data for testing a Monte-Carlo neutron and photon transport code
A Monte-Carlo photon and neutron transport code was developed at OAEP. The code was written in C and C++ languages in an object-oriented programming style. Constructive solid geometry (CSG), rather than combinatorial, was used such that making its input file more readable and recognizable. As the first stage of code validation, data from some ENDF files, in the MCNP's specific format, were used and compared with experimental data. The neutron (from a 300 mCi Am/Be source) attenuation by water was chosen to compare the results. The agreement of the quantity 1/Σ among the calculation from SIPHON and MCNP, and the experiment - which are 10.39 cm, 9.71 cm and 10.25 cm respectively - was satisfactorily well within the experimental uncertainties. These results also agree with the 10.8 cm result of N.M., Mirza, et al. (author)
3-D Monte Carlo neutron-photon transport code JMCT and its algorithms
JMCT Monte Carlo neutron and photon transport code has been developed which is based on the JCOGIN toolbox. JCOGIN includes the geometry operation, tally, the domain decomposition and the parallel computation about particle (MPI) and spatial domain (OpenMP) etc. The viewdata of CAD is equipped in JMCT preprocessor. The full-core pin-mode, which is from Chinese Qinshan-II nuclear power station, is design and simulated by JMCT. The detail pin-power distribution and keff results are shown in this paper. (author)
Two methods of calculating criticality are available in the 3D generalised geometry Monte Carlo particle transport code SPARTAN (Bending and Heffer, 1975). The first is a matrix technique in which the multiplication constant and source distribution of the system under study are calculated from estimates of fission probabilities and the second a method in which the multiplication constant is inferred from estimates of changes in neutron population over a number of neutron generations. Modifications are described which have been made to the way in which these methods are used in SPARTAN in order to improve the efficiency of criticality calculations. (author)
Pandya, Tara M.; Johnson, Seth R.; Evans, Thomas M.; Davidson, Gregory G.; Hamilton, Steven P.; Godfrey, Andrew T.
2016-03-01
This work discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package authored at Oak Ridge National Laboratory. Shift has been developed to scale well from laptops to small computing clusters to advanced supercomputers and includes features such as support for multiple geometry and physics engines, hybrid capabilities for variance reduction methods such as the Consistent Adjoint-Driven Importance Sampling methodology, advanced parallel decompositions, and tally methods optimized for scalability on supercomputing architectures. The scaling studies presented in this paper demonstrate good weak and strong scaling behavior for the implemented algorithms. Shift has also been validated and verified against various reactor physics benchmarks, including the Consortium for Advanced Simulation of Light Water Reactors' Virtual Environment for Reactor Analysis criticality test suite and several Westinghouse AP1000® problems presented in this paper. These benchmark results compare well to those from other contemporary Monte Carlo codes such as MCNP5 and KENO.
An object-oriented implementation of a parallel Monte Carlo code for radiation transport
Santos, Pedro Duarte; Lani, Andrea
2016-05-01
This paper describes the main features of a state-of-the-art Monte Carlo solver for radiation transport which has been implemented within COOLFluiD, a world-class open source object-oriented platform for scientific simulations. The Monte Carlo code makes use of efficient ray tracing algorithms (for 2D, axisymmetric and 3D arbitrary unstructured meshes) which are described in detail. The solver accuracy is first verified in testcases for which analytical solutions are available, then validated for a space re-entry flight experiment (i.e. FIRE II) for which comparisons against both experiments and reference numerical solutions are provided. Through the flexible design of the physical models, ray tracing and parallelization strategy (fully reusing the mesh decomposition inherited by the fluid simulator), the implementation was made efficient and reusable.
ACCEPT: three-dimensional electron/photon Monte Carlo transport code using combinatorial geometry
The ACCEPT code provides experimenters and theorists with a method for the routine solution of coupled electron/photon transport through three-dimensional multimaterial geometries described by the combinational method. Emphasis is placed upon operational simplicity without sacrificing the rigor of the model. ACCEPT combines condensed-history electron Monte Carlo with conventional single-scattering photon Monte Carlo in order to describe the transport of all generations of particles from several MeV down to 1.0 and 10.0 keV for electrons and photons, respectively. The model is more accurate at the higher energies with a less rigorous description of the particle cascade at energies where the shell structure of the transport media becomes important. Flexibility of construction permits the user to tailor the model to specific applications and to extend the capabilities of the model to more sophisticated applications through relatively simple update procedures. The ACCEPT code is currently running on the CDC-7600 (66000) where the bulk of the cross-section data and the statistical variables are stored in Large Core Memory
Point KENO V.a: A continuous-energy Monte Carlo code for transport applications
KENO V.a is a multigroup Monte Carlo code that solves the Boltzmann transport equation and is used extensively in the criticality safety community to calculate the effective multiplication factor of systems with fissionable material. In this work, a continuous-energy or pointwise version of KENO V.a has been developed by first designing a new continuous-energy cross-section format and then by developing the appropriate Monte Carlo transport procedures to sample the new cross-section format. In order to generate pointwise cross sections for a test library, a series of cross-section processing modules were developed and used to process 50 ENDF/B-VI Release 7 nuclides for the test library. Once the cross-section processing procedures were in place, a continuous-energy version of KENO V.a was developed and tested by calculating 46 test cases that include critical and calculational benchmark problems. The Point KENO-calculated results for the test problems are in agreement with calculated results obtained with the multigroup version of KENO V.a and MCNP4C. Based on the calculated results with the prototypic cross-section library, a continuous-energy version of the KENO V.a code has been successfully developed and demonstrated for modeling systems with fissionable material. (authors)
Nelson, Adam
Multi-group scattering moment matrices are critical to the solution of the multi-group form of the neutron transport equation, as they are responsible for describing the change in direction and energy of neutrons. These matrices, however, are difficult to correctly calculate from the measured nuclear data with both deterministic and stochastic methods. Calculating these parameters when using deterministic methods requires a set of assumptions which do not hold true in all conditions. These quantities can be calculated accurately with stochastic methods, however doing so is computationally expensive due to the poor efficiency of tallying scattering moment matrices. This work presents an improved method of obtaining multi-group scattering moment matrices from a Monte Carlo neutron transport code. This improved method of tallying the scattering moment matrices is based on recognizing that all of the outgoing particle information is known a priori and can be taken advantage of to increase the tallying efficiency (therefore reducing the uncertainty) of the stochastically integrated tallies. In this scheme, the complete outgoing probability distribution is tallied, supplying every one of the scattering moment matrices elements with its share of data. In addition to reducing the uncertainty, this method allows for the use of a track-length estimation process potentially offering even further improvement to the tallying efficiency. Unfortunately, to produce the needed distributions, the probability functions themselves must undergo an integration over the outgoing energy and scattering angle dimensions. This integration is too costly to perform during the Monte Carlo simulation itself and therefore must be performed in advance by way of a pre-processing code. The new method increases the information obtained from tally events and therefore has a significantly higher efficiency than the currently used techniques. The improved method has been implemented in a code system
Load balancing in highly parallel processing of Monte Carlo code for particle transport
In parallel processing of Monte Carlo(MC) codes for neutron, photon and electron transport problems, particle histories are assigned to processors making use of independency of the calculation for each particle. Although we can easily parallelize main part of a MC code by this method, it is necessary and practically difficult to optimize the code concerning load balancing in order to attain high speedup ratio in highly parallel processing. In fact, the speedup ratio in the case of 128 processors remains in nearly one hundred times when using the test bed for the performance evaluation. Through the parallel processing of the MCNP code, which is widely used in the nuclear field, it is shown that it is difficult to attain high performance by static load balancing in especially neutron transport problems, and a load balancing method, which dynamically changes the number of assigned particles minimizing the sum of the computational and communication costs, overcomes the difficulty, resulting in nearly fifteen percentage of reduction for execution time. (author)
A computer code package (PTSIM) for particle transport Monte Carlo simulation was developed using object oriented techniques of design and programming. A flexible system for simulation of coupled photon, electron transport, facilitating development of efficient simulation applications, was obtained. For photons: Compton and photo-electric effects, pair production and Rayleigh interactions are simulated, while for electrons, a class II condensed history scheme was considered, in which catastrophic interactions (Moeller electron-electron interaction, bremsstrahlung, etc.) are treated in detail and all other interactions with reduced individual effect on electron history are grouped together using continuous slowing down approximation and energy straggling theories. Electron angular straggling is simulated using Moliere theory or a mixed model in which scatters at large angles are treated as distinct events. Comparisons with experimentally benchmarks for electron transmission and bremsstrahlung emissions energy and angular spectra, and for dose calculations are presented
MCNP: a general Monte Carlo code for neutron and photon transport
The general-purpose Monte Carlo code MCNP ca be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation are accounted for. Thermal neutrons are described by both the free-gas and S(α,β) models. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. MCNP includes an elaborate, interactive plotting capability that allows the user to view his input geometry to help check for setup errors. Standard features which are available to improve computational efficiency include geometry splitting and Russian roulette, weight cutoff with Russian roulette, correlated sampling, analog capture or capture by weight reduction, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point or ring detectors, deterministically transporting pseudo-particles to designated regions, track-length estimators, source biasing, and several parameter cutoffs. Extensive summary information is provided to help the user better understand the physics and Monte Carlo simulation of his problem. The standard, user-defined output of MCNP includes two-way current as a function of direction across any set of surfaces or surface segments in the problem. Flux across any set of surfaces or surface segments is available. 58 figures, 28 tables
This paper summarized two improvements of a real production code by using vectorization and multitasking techniques. After a short description of Monte Carlo algorithms employed in our neutron transport problems, we briefly describe the work we have done in order to get a vector code. Vectorization principles will be presented and measured performances on the CRAY 1S, CYBER 205 and CRAY X-MP compared in terms of vector lengths. The second part of this work is an adaptation to multitasking on the CRAY X-MP using exclusively standard multitasking tools available with FORTRAN under the COS 1.13 system. Two examples will be presented. The goal of the first one is to measure the overhead inherent to multitasking when tasks become too small and to define a granularity threshold, that is to say a minimum size for a task. With the second example we propose a method that is very X-MP oriented in order to get the best speedup factor on such a computer. In conclusion we prove that Monte Carlo algorithms are very well suited to future vector and parallel computers. (orig.)
Comparison of Space Radiation Calculations from Deterministic and Monte Carlo Transport Codes
Adams, J. H.; Lin, Z. W.; Nasser, A. F.; Randeniya, S.; Tripathi, r. K.; Watts, J. W.; Yepes, P.
2010-01-01
The presentation outline includes motivation, radiation transport codes being considered, space radiation cases being considered, results for slab geometry, results from spherical geometry, and summary. ///////// main physics in radiation transport codes hzetrn uprop fluka geant4, slab geometry, spe, gcr,
MCPT: A Monte Carlo code for simulation of photon transport in tomographic scanners
MCPT is a special-purpose Monte Carlo code designed to simulate photon transport in tomographic scanners. Variance reduction schemes and sampling games present in MCPT were selected to characterize features common to most tomographic scanners. Combined splitting and biasing (CSB) games are used to systematically sample important detection pathways. An efficient splitting game is used to tally particle energy deposition in detection zones. The pulse height distribution of each detector can be found by convolving the calculated energy deposition distribution with the detector's resolution function. A general geometric modelling package, HERMETOR, is used to describe the geometry of the tomographic scanners and provide MCPT information needed for particle tracking. MCPT's modelling capabilites are described and preliminary experimental validation is presented. (orig.)
A reliable Monte Carlo based investigation of ion chambers in medical physics problems depends on the accuracy of the charged particle transport and implementations of the condensed history technique. Improper handling of media interfaces can lead to anomalous results or 'interface artefacts'. This work presents a rigorous investigation of the electron transport algorithm in the general purpose Monte Carlo (MC) code FLUKA (2008.3b.1). A 'Fano test' was implemented in order to benchmark the accuracy of the algorithm. Furthermore, the calculation of wall perturbation factors pwall of a Roos type chamber irradiated by electrons were performed and compared with values based on the EGSnrc MC code
TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files
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.
Tripoli-3: monte Carlo transport code for neutral particles - version 3.5 - users manual
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)
Verification of Monte Carlo transport codes: FLUKA, MARS and SHIELD-A
Monte Carlo transport codes like FLUKA, MARS and SHIELD are widely used for the estimation of radiation hazards in accelerator facilities. Accurate simulations are especially important with increasing energies and intensities of the machines. As the physical models implied in the codes are being constantly further developed, the verification is needed to make sure that the simulations give reasonable results. We report on the verification of electronic stopping modules and the verification of nuclide production modules of the codes. The verification of electronic stopping modules is based on the results of irradiation of stainless steel, copper and aluminum by 500 MeV/u and 950 MeV/u uranium ions. The stopping ranges achieved experimentally are compared with the simulated ones. The verification of isotope production modules is done via comparing the experimental depth profiles of residual activity (aluminum targets were irradiated by 500 MeV/u and 950 MeV/u uranium ions) with the results of simulations. Correspondences and discrepancies between the experiment and the simulations are discussed.
Verification of Monte Carlo transport codes: FLUKA, MARS and SHIELD-A
Chetvertkova, Vera [IAP, J. W. Goethe-University, Frankfurt am Main (Germany); GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Mustafin, Edil; Strasik, Ivan [GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Ratzinger, Ulrich [IAP, J. W. Goethe-University, Frankfurt am Main (Germany); Latysheva, Ludmila; Sobolevskiy, Nikolai [Institute for Nuclear Research RAS, Moscow (Russian Federation)
2011-07-01
Monte Carlo transport codes like FLUKA, MARS and SHIELD are widely used for the estimation of radiation hazards in accelerator facilities. Accurate simulations are especially important with increasing energies and intensities of the machines. As the physical models implied in the codes are being constantly further developed, the verification is needed to make sure that the simulations give reasonable results. We report on the verification of electronic stopping modules and the verification of nuclide production modules of the codes. The verification of electronic stopping modules is based on the results of irradiation of stainless steel, copper and aluminum by 500 MeV/u and 950 MeV/u uranium ions. The stopping ranges achieved experimentally are compared with the simulated ones. The verification of isotope production modules is done via comparing the experimental depth profiles of residual activity (aluminum targets were irradiated by 500 MeV/u and 950 MeV/u uranium ions) with the results of simulations. Correspondences and discrepancies between the experiment and the simulations are discussed.
A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom
Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H., E-mail: mbellezzo@gmail.br [Instituto de Pesquisas Energeticas e Nucleares / CNEN, Av. Lineu Prestes 2242, Cidade Universitaria, 05508-000 Sao Paulo (Brazil)
2014-08-15
As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)
A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom
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)
As The Monte Carlo (MC) particle transport analysis for a complex system such as research reactor, accelerator, and fusion facility may require accurate modeling of the complicated geometry. Its manual modeling by using the text interface of a MC code to define the geometrical objects is tedious, lengthy and error-prone. This problem can be overcome by taking advantage of modeling capability of the computer aided design (CAD) system. There have been two kinds of approaches to develop MC code systems utilizing the CAD data: the external format conversion and the CAD kernel imbedded MC simulation. The first approach includes several interfacing programs such as McCAD, MCAM, GEOMIT etc. which were developed to automatically convert the CAD data into the MCNP geometry input data. This approach makes the most of the existing MC codes without any modifications, but implies latent data inconsistency due to the difference of the geometry modeling system. In the second approach, a MC code utilizes the CAD data for the direct particle tracking or the conversion to an internal data structure of the constructive solid geometry (CSG) and/or boundary representation (B-rep) modeling with help of a CAD kernel. MCNP-BRL and OiNC have demonstrated their capabilities of the CAD-based MC simulations. Recently we have developed a CAD-based geometry processing module for the MC particle simulation by using the OpenCASCADE (OCC) library. In the developed module, CAD data can be used for the particle tracking through primitive CAD surfaces (hereafter the CAD-based tracking) or the internal conversion to the CSG data structure. In this paper, the performances of the text-based model, the CAD-based tracking, and the internal CSG conversion are compared by using an in-house MC code, McSIM, equipped with the developed CAD-based geometry processing module
Aurora T: a Monte Carlo code for transportation of neutral atoms in a toroidal plasma
This paper contains a short description of Aurora code. This code have been developed at Princeton with Monte Carlo method for calculating neutral gas in cylindrical plasma. In this work subroutines such one can take in account toroidal geometry are developed
MCNP: a general Monte Carlo code for neutron and photon transport. Version 3A. Revision 2
This manual is a practical guide for the use of our general-purpose Monte Carlo code MCNP. The first chapter is a primer for the novice user. The second chapter describes the mathematics, data, physics, and Monte Carlo simulation found in MCNP. This discussion is not meant to be exhaustive - details of the particular techniques and of the Monte Carlo method itself will have to be found elsewhere. The third chapter shows the user how to prepare input for the code. The fourth chapter contains several examples, and the fifth chapter explains the output. The appendices show how to use MCNP on particular computer systems at the Los Alamos National Laboratory and also give details about some of the code internals that those who wish to modify the code may find useful. 57 refs
Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code
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)
MCNP: a general Monte Carlo code for neutron and photon transport
The general-purpose Monte Carlo code MCNP can be used for neutron, photon, or coupled neutron--photon transport. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-IV) are accounted for. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. Standard optional variance reduction schemes include geometry splitting and Russian roulette, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point detectors, track-length estimators, and source biasing. The standard output of MCNP includes two-way current as a function of energy, time, and angle with the normal, across any subset of bounding surfaces in the problem. Fluxes across any set of bounding surfaces are available as a function of time and energy. Similarly, the flux at designated points and the average flux in a cell (track length per unit volume) are standard tallies. Reactions such as fissions or absorptions may be obtained in a subset of geometric cells. The heating tallies give the energy deposition per starting particle. In addition, particles may be flagged when they cross specified surfaces or enter designated cells, and the contributions of these flagged particles to certain of the tallies are listed separately. All quantities printed out have their relative errors listed also. 11 figures, 27 tables
Modelling of a general purpose irradiation chamber using a Monte Carlo particle transport code
Full-text: The aim of this research is to stimulate the effectiveness use of a general purpose irradiation chamber to contain pure neutron particles obtained from a research reactor. The secondary neutron and gamma particles dose discharge from the chamber layers will be used as a platform to estimate the safe dimension of the chamber. The chamber, made up of layers of lead (Pb), shielding, polyethylene (PE), moderator and commercial grade aluminium (Al) cladding is proposed for the use of interacting samples with pure neutron particles in a nuclear reactor environment. The estimation was accomplished through simulation based on general Monte Carlo N-Particle transport code using Los Alamos MCNPX software. Simulations were performed on the model of the chamber subjected to high neutron flux radiation and its gamma radiation product. The model of neutron particle used is based on the neutron source found in PUSPATI TRIGA MARK II research reactor which holds a maximum flux value of 1 x 1012 neutron/ cm2s. The expected outcomes of this research are zero gamma dose in the core of the chamber and neutron dose rate of less than 10 μSv/ day discharge from the chamber system. (author)
Memory bottlenecks and memory contention in multi-core Monte Carlo transport codes
Highlights: • The performance of nuclear reactor Monte Carlo transport applications is examined. • A “proxy-application” (XSBench) is presented representing the key kernel. • In-depth performance analyses reveal the algorithm is bottlenecked by bandwidth. • Strategies are discussed to improve scalability on next generation HPC systems. - Abstract: We have extracted a kernel that executes only the most computationally expensive steps of the Monte Carlo particle transport algorithm – the calculation of macroscopic cross sections – in an effort to expose bottlenecks within multi-core, shared memory architectures
Its version 3.0. The integrated TIGER series of coupled electron/photon Monte Carlo transport codes
The ITS system is described, which is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields. It combines operational simplicity and physical accuracy in order to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Flexibility of construction permits tailoring of the codes to specific applications and extension of code capabilities to more complex applications through simple update procedures. (author)
Monte-Carlo Impurity transport simulations in the edge of the DIII-D tokamak using the MCI code
A Monte-Carlo Impurity (MCI) transport code is used to follow trace impurities through multiple ionization states in realistic 2-D tokamak geometries. The MCI code is used to study impurity transport along the open magnetic field lines of the Scrape-off Layer (SOL) and to understand how impurities get into the core from the SOL. An MCI study concentrating on the entrainment of carbon impurities ions by deuterium background plasma into the DIII-D divertor is discussed. MCI simulation results are compared to experimental DIII-D carbon measurements
Development of a GPU-based Monte Carlo dose calculation code for coupled electron-photon transport
Jia, Xun; Sempau, Josep; Choi, Dongju; Majumdar, Amitava; Jiang, Steve B
2009-01-01
Monte Carlo simulation is the most accurate method for absorbed dose calculations in radiotherapy. Its efficiency still requires improvement for routine clinical applications, especially for online adaptive radiotherapy. In this paper, we report our recent development on a GPU-based Monte Carlo dose calculation code for coupled electron-photon transport. We have implemented the Dose Planning Method (DPM) Monte Carlo dose calculation package (Sempau et al, Phys. Med. Biol., 45(2000)2263-2291) on GPU architecture under CUDA platform. The implementation has been tested with respect to the original sequential DPM code on CPU in two cases. Our results demonstrate the adequate accuracy of the GPU implementation for both electron and photon beams in radiotherapy energy range. A speed up factor of 4.5 and 5.5 times have been observed for electron and photon testing cases, respectively, using an NVIDIA Tesla C1060 GPU card against a 2.27GHz Intel Xeon CPU processor .
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.
ITS version 5.0 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes
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
Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code
High-temperature gas-cooled reactor fuels are composed of thousands of microscopic fuel particles, randomly dispersed in a graphite matrix. The modelling of such geometry is complicated, especially using continuous-energy Monte Carlo codes, which are unable to apply any deterministic corrections in the calculation. This paper presents the geometry routine developed for modelling randomly dispersed particle fuels using the PSG Monte Carlo reactor physics code. The model is based on the delta-tracking method, and it takes into account the spatial self-shielding effects and the random dispersion of the fuel particles. The calculation routine is validated by comparing the results to reference MCNP4C calculations using uranium and plutonium based fuels. (authors)
Vectorization and multitasking with a Monte-Carlo code for neutron transport problems
This paper summarizes two improvements of a Monte Carlo code by resorting to vectorization and multitasking techniques. After a short presentation of the physical problem to solve and a description of the main difficulties to produce an efficient coding, this paper introduces the vectorization principles employed and briefly describes how the vectorized algorithm works. Next, measured performances on CRAY 1S, CYBER 205 and CRAY X-MP are compared. The second part of this paper is devoted to multitasking technique. Starting from the standard multitasking tools available with FORTRAN on CRAY X-MP/4, a multitasked algorithm and its measured speed-ups are presented. In conclusion we prove that vector and parallel computers are a great opportunity for such Monte Carlo algorithms
SPHERE: a spherical-geometry multimaterial electron/photon Monte Carlo transport code
SPHERE provides experimenters and theorists with a method for the routine solution of coupled electron/photon transport through multimaterial configurations possessing spherical symmetry. Emphasis is placed upon operational simplicity without sacrificing the rigor of the model. SPHERE combines condensed-history electron Monte Carlo with conventional single-scattering photon Monte Carlo in order to describe the transport of all generations of particles from several MeV down to 1.0 and 10.0 keV for electrons and photons, respectively. The model is more accurate at the higher energies, with a less rigorous description of the particle cascade at energies where the shell structure of the transport media becomes important. Flexibility of construction permits the user to tailor the model to specific applications and to extend the capabilities of the model to more sophisticated applications through relatively simple update procedures. 8 figs., 3 tables
Exact modeling of the torus geometry with Monte Carlo transport code
It is valuable to model torus geometry exactry for the neutronics design of fusion reactor in order to assess neutronics characteristics such as tritium breeding ratio, heat generation rate, etc, near the plasma. Monte Carlo code MORSE-GG which plays important role in the radiation streaming calculation of fusion reactors had been able to deal with the geometry composed of second order surfaces. The MORSE-GG program is modified to be able to deal with torus geometry which has fourth order surface by solving biquadratic equations, hoping that MORSE-GG code becomes more effective for the neutronics calculation of the Tokamak fusion reactor. (author)
Installation of Monte Carlo neutron and photon transport code system MCNP4
The continuous energy Monte Carlo code MCNP-4 including its graphic functions has been installed on the Sun-4 sparc-2 work station with minor corrections. In order to validate the installed MCNP-4 code, 25 sample problems have been executed on the work station and these results have been compared with the original ones. And, the most of the graphic functions have been demonstrated by using 3 sample problems. Further, additional 14 nuclides have been included to the continuous cross section library edited from JENDL-3. (author)
A Monte Carlo Code for Relativistic Radiation Transport Around Kerr Black Holes
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.
Antiproton annihilation physics in the Monte Carlo particle transport code SHIELD-HIT12A
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
Realistic simulations of the passage of fast neutrons through tissue require a large quantity of cross-sectional data. What are needed are differential (in particle type, energy and angle) cross sections. A computer code is described which produces such spectra for neutrons above ∼14 MeV incident on light nuclei such as carbon and oxygen. Comparisons have been made with experimental measurements of double-differential secondary charged-particle production on carbon and oxygen at energies from 27 to 60 MeV; they indicate that the model is adequate in this energy range. In order to utilize fully the results of these calculations, they should be incorporated into a neutron transport code. This requires defining a generalized format for describing charged-particle production, putting the calculated results in this format, interfacing the neutron transport code with these data, and charged-particle transport. The design and development of such a program is described. 13 refs., 3 figs
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.
BOT3P consists of a set of standard Fortran 77 language programs that gives the users of the deterministic transport codes DORT, TORT, TWODANT, THREEDANT, PARTISN and the sensitivity code SUSD3D some useful diagnostic tools to prepare and check the geometry of their input data files for both Cartesian and cylindrical geometries, including graphical display modules. Users can produce the geometrical and material distribution data for all the cited codes for both two-dimensional and three-dimensional applications and, only in 3-dimensional Cartesian geometry, for the Monte Carlo Transport Code MCNP, starting from the same BOT3P input. Moreover, BOT3P stores the fine mesh arrays and the material zone map in a binary file, the content of which can be easily interfaced to any deterministic and Monte Carlo transport code. This makes it possible to compare directly for the same geometry the effects stemming from the use of different data libraries and solution approaches on transport analysis results. BOT3P Version 5.0 lets users optionally and with the desired precision compute the area/volume error of material zones with respect to the theoretical values, if any, because of the stair-cased representation of the geometry, and automatically update material densities on the whole zone domains to conserve masses. A local (per mesh) density correction approach is also available. BOT3P is designed to run on Linux/UNIX platforms and is publicly available from the Organization for Economic Cooperation and Development (OECD/NEA)/Nuclear Energy Agency Data Bank. Through the use of BOT3P, radiation transport problems with complex 3-dimensional geometrical structures can be modelled easily, as a relatively small amount of engineer-time is required and refinement is achieved by changing few parameters. This tool is useful for solving very large challenging problems, as successfully demonstrated not only in some complex neutron shielding and criticality benchmarks but also in a power
The Premar Code for the Monte Carlo Simulation of Radiation Transport In the Atmosphere
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
Domain decomposition and terabyte tallies with the OpenMC Monte Carlo neutron transport code
Memory limitations are a key obstacle to applying Monte Carlo neutron transport methods to high-fidelity full-core reactor analysis. Billions of unique regions are needed to carry out full-core depletion and fuel performance analyses, equating to terabytes of memory for isotopic abundances and tally scores - far more than can fit on a single computational node in modern architectures. This work introduces an implementation of domain decomposition that addresses this problem, demonstrating excellent scaling up to a 2.39TB mesh-tally distributed across 512 compute nodes running a full-core reactor benchmark on the Mira Blue Gene/Q supercomputer at Argonne National Laboratory. (author)
Radiation transport in random disperse media implemented in the Monte Carlo code PRIZMA
The paper describes PRIZMA capabilities for modeling radiation transport in random disperse media by the Monte Carlo method. It proposes a method for simulating radiation transport in binary media with variable volume fractions. The method models the medium consequently from one grain crossed by a particle trajectory to another. Like in the Limited Chord Length Sampling (LCLS) method, particles in grains are tracked in the actual grain geometry, but unlike LCLS, the medium is modeled using only Matrix Chord Length Sampling (MCLS) from the exponential distribution and it is not necessary to know the grain chord length distribution. This helped us extend the method to media with randomly oriented, arbitrarily shaped convex grains. Other extensions include multicomponent media - grains of several sorts, and polydisperse media - grains of different sizes
The Monte Carlo method was used to build a new code for the simulation of particle transport. Several calculations were done after that for verification, where different sources were used, the source term was obtained using the ORIGEN-S code. Water and lead shield were used with spherical geometry, and the tally results were obtained on the external surface of the shield, afterward the results were compared with the results of MCNPX for verification of the new code. The variance reduction techniques of splitting and Russian Roulette were implemented in the code to be more efficient, by reducing the amount of custom programming required, by artificially increasing the particles being tallied with decreasing the weight. The code shows lower results than the results of MCNPX, this can be interpreted by the effect of the secondary gamma radiation that can be produced by the electron, which is ejected by the primary radiation. In the future a more study will be made on the effect of the electron production and transport, either by a real transport of the electron or by simply using an approximation such the thick target bremsstahlung(TTB) option which is used in MCNPX
Alnajjar, Alaaddin [Univ. of Science and Technology, Daejeon (Korea, Republic of); Park, Chang Je; Lee, Byunchul [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2013-10-15
The Monte Carlo method was used to build a new code for the simulation of particle transport. Several calculations were done after that for verification, where different sources were used, the source term was obtained using the ORIGEN-S code. Water and lead shield were used with spherical geometry, and the tally results were obtained on the external surface of the shield, afterward the results were compared with the results of MCNPX for verification of the new code. The variance reduction techniques of splitting and Russian Roulette were implemented in the code to be more efficient, by reducing the amount of custom programming required, by artificially increasing the particles being tallied with decreasing the weight. The code shows lower results than the results of MCNPX, this can be interpreted by the effect of the secondary gamma radiation that can be produced by the electron, which is ejected by the primary radiation. In the future a more study will be made on the effect of the electron production and transport, either by a real transport of the electron or by simply using an approximation such the thick target bremsstahlung(TTB) option which is used in MCNPX.
The aim of this study was to investigate the Monte Carlo (MC) code FLUKA, regarding its ability to accurately simulate electron transport at density inhomogeneities and in ionization chamber geometries. In order to evaluate the accuracy of FLUKA's electron transport algorithm and the implementation of the condensed history technique, a Fano test was used. This test allows the comparison of calculated and theoretically expected results. The ratio of the two results is ideally equal to unity, and a deviation usually indicates artifacts in the treatment of density interfaces. As a more practical problem, wall perturbation factors pwall of a plane parallel chamber in electron beams were calculated and compared with results based on the EGSnrc MC code. Additionally, the impact of wall material and thickness on calculated cavity dose was investigated for two different thimble chambers irradiated by 60Co. The correct choice of parameters within FLUKA's electron transport algorithm ensured passing the Fano test within ∼0.7% and a good agreement for practical examples within 0.4% compared to results of the EGSnrc MC code. The latter is known to allow an artifact free simulation of ionization chamber response in photon and electron beams. Based on these results, the electron transport accuracy within the FLUKA code can generally be regarded as much better than 1% for typical ionization chamber dosimetry problems. (author)
A Monte Carlo transport code study of the space radiation environment using FLUKA and ROOT
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...
Towards scalable parellelism in Monte Carlo particle transport codes using remote memory access
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.
Stochastic-media simulations require numerous boundary crossings. We consider two Monte Carlo electron transport approaches and evaluate accuracy with numerous material boundaries. In the condensed-history method, approximations are made based on infinite-medium solutions for multiple scattering over some track length. Typically, further approximations are employed for material-boundary crossings where infinite-medium solutions become invalid. We have previously explored an alternative 'condensed transport' formulation, a Generalized Boltzmann-Fokker-Planck (GBFP) method, which requires no special boundary treatment but instead uses approximations to the electron-scattering cross sections. Some limited capabilities for analog transport and a GBFP method have been implemented in the Integrated Tiger Series (ITS) codes. Improvements have been made to the condensed history algorithm. The performance of the ITS condensed-history and condensed-transport algorithms are assessed for material-boundary crossings. These assessments are made both by introducing artificial material boundaries and by comparison to analog Monte Carlo simulations. (authors)
Geometry system used in the General Monte Carlo transport code SPARTAN
The geometry routines used in the general-purpose, three-dimensional particle transport code SPARTAN are described. The code is designed to deal with the very complex geometries encountered in lattice cell and fuel handling calculations, health physics, and shielding problems. Regions of the system being studied may be represented by simple shapes (spheres, cylinders, and so on) or by multinomial surfaces of any order, and many simple shapes may be combined to make up a complex layout. The geometry routines are designed to allow the program to carry out a number of tasks (such as sampling for a random point or tracking a path through several regions) in any order, so that the use of the routines is not restricted to a particular tracking or scoring method. Routines for reading, checking, and printing the data are included. (U.S.)
Modern Monte Carlo transport simulations of the Lawrence Livermore National Laboratory pulsed-sphere time of flight experiments have recently been performed. In these experiments, 14 MeV neutrons, generated via the 3H(d, n)4He reaction, interact with a sphere of material that surrounds the neutron generating target. The time of arrival of the uncollided and collided neutrons are recorded in a detector system placed up to 10 meters from the center of the sphere. A collection of experiments with varying sphere materials, mean-free-paths and detector systems have been modeled using the Mercury Monte Carlo transport code. This effort serves to validate new features of the Mercury code, including general sources, tallies and point-detector / biased-collisions variance reduction methods, as well as assess the quality of evaluated nuclear data sets. In general, the level of agreement between the calculations and experiment is very good. However, for certain pulsed spheres, discrepancies are observed between the simulations using different nuclear data sets. (author)
In order to improve the accuracy and calculating speed of shielding analyses, MCNP 4, a Monte Carlo neutron and photon transport code system, has been parallelized and measured of its efficiency in the highly parallel distributed memory type computer, AP1000. The code has been analyzed statically and dynamically, then the suitable algorithm for parallelization has been determined for the shielding analysis functions of MCNP 4. This includes a strategy where a new history is assigned to the idling processor element dynamically during the execution. Furthermore, to avoid the congestion of communicative processing, the batch concept, processing multi-histories by a unit, has been introduced. By analyzing a sample cask problem with 2,000,000 histories by the AP1000 with 512 processor elements, the 82 % of parallelization efficiency is achieved, and the calculational speed has been estimated to be around 50 times as fast as that of FACOM M-780. (author)
The geometry system of the three-dimensional Monte Carlo particle transport code SPARTAN
The geometry routines used in the general-purpose three-dimensional particle transport code SPARTAN are described. The code is designed to deal with the very complex geometries encountered in lattice cell and fuel handling calculations, health physics and shielding problems. Regions of the system being studied may be represented by simple shapes (spheres, cylinders etc) or by multinomial surfaces of any order, and many simple shapes may be combined to make up a complex layout. The geometry routines are designed to allow the program to carry out a number of tasks (such as sampling for a random point or tracking a path through several regions) in any order, so that the use of the routines is not restricted to a particular tracking of scoring method. Routines for reading, checking and printing the data are included. Details of the computational package are also included to indicate the way in which the generalised geometry capability of SPARTAN could be incorporated into other codes. (author)
User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code
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 keff (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)
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
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)
Greenman, G M; O' Brien, M J; Procassini, R J; Joy, K I
2009-03-09
Two enhancements to the combinatorial geometry (CG) particle tracker in the Mercury Monte Carlo transport code are presented. The first enhancement is a hybrid particle tracker wherein a mesh region is embedded within a CG region. This method permits efficient calculations of problems with contain both large-scale heterogeneous and homogeneous regions. The second enhancement relates to the addition of parallelism within the CG tracker via spatial domain decomposition. This permits calculations of problems with a large degree of geometric complexity, which are not possible through particle parallelism alone. In this method, the cells are decomposed across processors and a particles is communicated to an adjacent processor when it tracks to an interprocessor boundary. Applications that demonstrate the efficacy of these new methods are presented.
Nowadays, radioactive isotopes are used in many different fields, for instance in industry, energy production, archaeology and mainly in medical applications. In addition, bricks and stones, which are used to build these buildings and our homes, have higher natural radiation levels than other building materials such as wood. In this work, the linear and mass attenuation coefficients of different types building materials, needed for the protection of human health against radiation hazards, were investigated with Monte Carlo particle-transport code (MCNP) technique. Simulations were performed in order to obtain these coefficients at photon energies from 80 keV to 1333 keV for clay, perlite and PP. As should be anticipated, the density and photon energy are the main parameters that affect the mass attenuation coefficient
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
Monte Carlo simulation code on impurity transport has been developed by several groups to be utilized mainly for fusion related edge plasmas. State of impurity particle is determined by atomic and molecular processes such as ionization, charge exchange in plasma. A lot of atomic and molecular processes have been considered because the edge plasma has not only impurity atoms, but also impurity molecules mainly related to chemical erosion of carbon materials, and their cross sections have been given experimentally and theoretically. We need to reveal which process is essential in a given edge plasma condition. Monte Carlo simulation code, which takes such various atomic and molecular processes into account, is necessary to investigate the behavior of impurity particle in plasmas. Usually, the impurity transport simulation code has been intended for some specific atomic and molecular processes so that the introduction of a new process forces complicated programming work. In order to evaluate various proposed atomic and molecular processes, a flexible management of atomic and molecular reaction should be established. We have developed the impurity transport simulation code based on object-oriented method. By employing object-oriented programming, we can handle each particle as 'object', which enfolds data and procedure function itself. A user (notice, not programmer) can define property of each particle species and the related atomic and molecular processes and then each 'object' is defined by analyzing this information. According to the relation among plasma particle species, objects are connected with each other and change their state by themselves. Dynamic allocation of these objects to program memory is employed to adapt for arbitrary number of species and atomic/molecular reactions. Thus we can treat arbitrary species and process starting from, for instance, methane and acetylene. Such a software procedure would be useful also for industrial application plasmas
The single interaction Monte Carlo code TRAX has been extended to describe low-energy electron creation and transport in solids. Electrons with energies below 1 keV have ranges in solids on the nanometerscale. Complete sets of electron interaction cross sections for energies below 1 keV down to 1 eV have been compiled and assessed for various target materials. The applicability of the cross sections has been validated by comparisons with experimental data as far as available. The code has further been extended to handle the production of Auger electrons and cascades. Furthermore, the capability to handle non-uniform targets has been added. With the extended TRAX code, experimental data from GSI's Toroid electron spectrometer have been reproduced using thin solid state foils of carbon, nickel, silver and gold as targets. Furthermore, the radial dose distribution around ion tracks has been investigated on the nanometer scale. The explicit consideration of Auger electron cascades has been used to evaluate whether metallic nanoparticles can locally enhance the dose in combination with proton or electron irradiation.
Successful vectorization - reactor physics Monte Carlo code
Most particle transport Monte Carlo codes in use today are based on the ''history-based'' algorithm, wherein one particle history at a time is simulated. Unfortunately, the ''history-based'' approach (present in all Monte Carlo codes until recent years) is inherently scalar and cannot be vectorized. In particular, the history-based algorithm cannot take advantage of vector architectures, which characterize the largest and fastest computers at the current time, vector supercomputers such as the Cray X/MP or IBM 3090/600. However, substantial progress has been made in recent years in developing and implementing a vectorized Monte Carlo algorithm. This algorithm follows portions of many particle histories at the same time and forms the basis for all successful vectorized Monte Carlo codes that are in use today. This paper describes the basic vectorized algorithm along with descriptions of several variations that have been developed by different researchers for specific applications. These applications have been mainly in the areas of neutron transport in nuclear reactor and shielding analysis and photon transport in fusion plasmas. The relative merits of the various approach schemes will be discussed and the present status of known vectorization efforts will be summarized along with available timing results, including results from the successful vectorization of 3-D general geometry, continuous energy Monte Carlo. (orig.)
The ITS (Integrated Tiger Series) Monte Carlo code package developed at Sandia National Laboratories and distributed as CCC-467/ITS by the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory (ORNL) consists of eight codes - the standard codes, TIGER, CYLTRAN, ACCEPT; the P-codes, TIGERP, CYLTRANP, ACCEPTP; and the M-codes ACCEPTM, CYLTRANM. The codes have been adapted to run on the IBM 3081, VAX 11/780, CDC-7600, and Cray 1 with the use of the update emulator UPEML. This manual should serve as a guide to a user running the codes on IBM computers having 370 architecture. The cases listed were tested on the IBM 3033, under the MVS operating system using the VS Fortran Level 1.3.1 compiler
Consistency test of the electron transport algorithm in the GEANT4 Monte Carlo code
In this work, the condensed history algorithm in GEANT4 (version 4.6.2.p01) is examined. We performed simulations of an ionization chamber composed of water for 1.25 MeV incident photon beams under Fano conditions, and evaluated the consistency of the cavity response for several combinations of electron transport parameters. GEANT4 permits electrons to reach geometric boundaries in large steps, and underestimates lateral displacement near interfaces. Step size artefacts due to distortions in electron fluence and angular distributions reduce the cavity dose by up to 39%. Accurate cavity response can be achieved using severe user-imposed step size restrictions. We suggest that improvements in the electron transport algorithm in GEANT4 should address the handling of boundary crossing
Modeling of tungsten transport in the linear plasma device PSI-2 with the 3D Monte-Carlo code ERO
Marenkov, E.; Eksaeva, A.; Borodin, D.; Kirschner, A.; Laengner, M.; Kurnaev, V.; Kreter, A.; Coenen, J. W.; Rasinski, M.
2015-08-01
The ERO code was modified for modeling of plasma-surface interactions and impurities transport in the PSI-2 installation. Results of experiments on tungsten target irradiation with argon plasma were taken as a benchmark for the new version of the code. Spectroscopy data modeled with the code are in good agreement with experimental ones. Main factors contributing to observed discrepancies are discussed.
The crucial problem for radiation shielding design at heavy ion accelerator facilities with beam energies of several GeV/n is the source term problem. Experimental data on double differential neutron yields from thick targets irradiated with high-energy uranium nuclei are lacking. At present there are not many Monte Carlo multipurpose codes that can work with primary high-energy uranium nuclei. These codes use different physical models for simulating nucleus-nucleus reactions. Therefore, verification of the codes with available experimental data is very important for selection of the most reliable code for practical tasks. This paper presents comparisons of the FLUKA, GEANT4 and SHIELD code simulations with experimental data on neutron production at 1 GeV/n 238U beam interaction with a thick Fe target
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V and V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to
Morgan C. White
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
Guideline for radiation transport simulation with the Monte Carlo method
Today, the photon and neutron transport calculations with the Monte Carlo method have been progressed with advanced Monte Carlo codes and high-speed computers. Monte Carlo simulation is rather suitable expression than the calculation. Once Monte Carlo codes become more friendly and performance of computer progresses, most of the shielding problems will be solved by using the Monte Carlo codes and high-speed computers. As those codes prepare the standard input data for some problems, the essential techniques for solving the Monte Carlo method and variance reduction techniques of the Monte Carlo calculation might lose the interests to the general Monte Carlo users. In this paper, essential techniques of the Monte Carlo method and the variance reduction techniques, such as importance sampling method, selection of estimator, and biasing technique, are described to afford a better understanding of the Monte Carlo method and Monte Carlo code. (author)
Common misconceptions in Monte Carlo particle transport
Booth, Thomas E., E-mail: teb@lanl.gov [LANL, XCP-7, MS F663, Los Alamos, NM 87545 (United States)
2012-07-15
Monte Carlo particle transport is often introduced primarily as a method to solve linear integral equations such as the Boltzmann transport equation. This paper discusses some common misconceptions about Monte Carlo methods that are often associated with an equation-based focus. Many of the misconceptions apply directly to standard Monte Carlo codes such as MCNP and some are worth noting so that one does not unnecessarily restrict future methods. - Highlights: Black-Right-Pointing-Pointer Adjoint variety and use from a Monte Carlo perspective. Black-Right-Pointing-Pointer Misconceptions and preconceived notions about statistical weight. Black-Right-Pointing-Pointer Reasons that an adjoint based weight window sometimes works well or does not. Black-Right-Pointing-Pointer Pulse height/probability of initiation tallies and 'the' transport equation. Black-Right-Pointing-Pointer Highlights unnecessary preconceived notions about Monte Carlo transport.
The Monte Carlo code MONK is a general program written to provide a high degree of flexibility to the user. MONK is distinguished by its detailed representation of nuclear data in point form i.e., the cross-section is tabulated at specific energies instead of the more usual group representation. The nuclear data are unadjusted in the point form but recently the code has been modified to accept adjusted group data as used in fast and thermal reactor applications. The various geometrical handling capabilities and importance sampling techniques are described. In addition to the nuclear data aspects, the following features are also described; geometrical handling routines, tracking cycles, neutron source and output facilities. 12 references. (U.S.)
calculation solver SNATCH in the PARIS code platform. The latter uses the transport theory which is indispensable for the new generation fast reactors analysis. The principal conclusions are as follows: The Monte-Carlo assembly calculation code is an interesting way (in the sense of avoiding the difficulties in the self-shielding calculation, the limited order development of anisotropy parameters, the exact 3D geometries) to validate the deterministic codes like ECCO or APOLLO3 and to produce the multi-group constants for deterministic or Monte-Carlo multi-group calculation codes. The results obtained for the moment with the multi-group constants calculated by TRIPOLI-4 code are comparable with those produced from ECCO, but did not show remarkable advantages. (author)
In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector super-computers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, (5) continuous-energy calculation at arbitrary temperatures, (6) estimation of real variances in eigenvalue problems, (7) point detector and surface crossing estimators, (8) statistical geometry model, (9) function of reactor noise analysis (simulation of the Feynman-α experiment), (10) arbitrary shaped lattice boundary, (11) periodic boundary condition, (12) parallelization with standard libraries (MPI, PVM), (13) supporting many platforms, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them. (author)
The SHIELD Monte-Carlo transport code [1-2] simulates the interactions of hadrons and atomic nuclei of arbitrary charge and mass number (Z,A) with complex extended targets in a wide energy range, from 1 TeV/u down to 1 MeV/u or thermal energies in the case of neutrons. SHIELD is used for solving the same type of problems as other well-known Monte-Carlo codes, e.g. LAHET, HERMES, FLUKA, GEANT or MCNPX. Nuclear reactions are taken into account in an exclusive approach where all stages of hadron nucleus and nucleus-nucleus interactions are described. Thus, SHIELD can be used to simulate interactions of heavy ions with complex macroscopic targets. The version of the SHIELD code extended to simulate heavy ions transport in the context of the beam therapy in oncology is called SHIELD-HIT (Heavy Ion Transport) [3]. The results presented in this work stem from a cooperative research project between the Department of Medical Radiation Physics, Karolinska Institute (Stockholm) and the Institute for Nuclear Research of the Russian Academy of Science (Moscow). (author)
Monte Carlo electron/photon transport
A review of nonplasma coupled electron/photon transport using Monte Carlo method is presented. Remarks are mainly restricted to linerarized formalisms at electron energies from 1 keV to 1000 MeV. Applications involving pulse-height estimation, transport in external magnetic fields, and optical Cerenkov production are discussed to underscore the importance of this branch of computational physics. Advances in electron multigroup cross-section generation is reported, and its impact on future code development assessed. Progress toward the transformation of MCNP into a generalized neutral/charged-particle Monte Carlo code is described. 48 refs
SERPENT Monte Carlo reactor physics code
SERPENT is a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, developed at VTT Technical Research Centre of Finland since 2004. The code is specialized in lattice physics applications, but the universe-based geometry description allows transport simulation to be carried out in complicated three-dimensional geometries as well. The suggested applications of SERPENT include generation of homogenized multi-group constants for deterministic reactor simulator calculations, fuel cycle studies involving detailed assembly-level burnup calculations, validation of deterministic lattice transport codes, research reactor applications, educational purposes and demonstration of reactor physics phenomena. The Serpent code has been publicly distributed by the OECD/NEA Data Bank since May 2009 and RSICC in the U. S. since March 2010. The code is being used in some 35 organizations in 20 countries around the world. This paper presents an overview of the methods and capabilities of the Serpent code, with examples in the modelling of WWER-440 reactor physics. (Author)
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
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.
Highlights: • Overview of the capabilities and features of the MC21 Monte Carlo code, version 6. • Detailed description of in-line reactor feedback capabilities in MC21. • Discussion of running strategies for Monte Carlo simulations with feedback effects. • Includes representative MC21 results for massively-parallel 3D reactor simulations. - Abstract: 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
DIANE multiparticle transport code
DIANE is a three-dimensional multigroup Monte Carlo code that performs coupled neutron-photon-electron and light charged particle Monte Carlo transport on a fixed geometry and/or an imploding material (for example an Inertial-Confinement- Fusion plasma target). DIANE's multigroup data are provided by the nuclear data code ZADIG. ZADIG groups cross sections from international data libraries in ENDF7 format. In order to better account for resonances, a specific multigroup treatment, called 'mini-group' has been implemented. For photons, photo-atomic interactions are considered as well as photonuclear reactions. Coherent scattering may include anomalous scattering factors. Incoherent scattering include scattering function from EPDL97. We also account for fluorescence X-rays and Auger electrons. For neutrons, all the reaction channels relevant are processed by ZADIG. The ZADIG code integrates the Secondary Source Bremsstrahlung (SSB) model that simulates electron-photon cascade from electron slowing down. For light ions, DIANE solves the Boltzmann-Fokker-Planck equation. The continuous effect of Coulomb interaction with the background plasma can be treated by three Fokker- Planck terms (slowing-down, energy straggling and angular deflection). The effect of electric and magnetic fields through the Lorentz force can also be taken into account
In this report, the results of KALIMER (Korea Advanced LIquid MEtal Reactor) core design calculated by the K-CORE computing system are compared and analyzed with those of MCDEP calculation. The effective multiplication factor, flux distribution, fission power distribution and the number densities of the important nuclides effected from the depletion calculation for the R-Z model and Hex-Z model of KALIMER core are compared. It is confirmed that the results of K-CORE system compared with those of MCDEP based on the Monte Carlo transport theory method agree well within 700 pcm for the effective multiplication factor estimation and also within 2% in the driver fuel region, within 10% in the radial blanket region for the reaction rate and the fission power density. Thus, the K-CORE system for the core design of KALIMER by treating the lumped fission product and mainly important nuclides can be used as a core design tool keeping the necessary accuracy
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)
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 optimization. PMID:24600168
Monte Carlo simulation code modernization
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...
Monte Carlo photon transport techniques
The basis of Monte Carlo calculation of photon transport problems is the computer simulation of individual photon histories and their subsequent averaging to provide the quantities of interest. As the history of a photon is followed the values of variables are selected and decisions made by sampling known distributions using random numbers. The transport of photon is simulated by creation of particles from a defined source region, generally with a random initial orientation in space, with tracking of particles as they travel through the system, sampling the probability density functions for their interactions to evaluate their trajectories and energy deposition at different points in the system. The interactions determine the penetration and the motion of particles. The computational model, for radiation transport problems includes geometry and material specifications. Every computer code contains a database of experimentally obtained quantities, known as cross-sections that determine the probability of a particle interacting with the medium through which it is transported. Every cross-section is peculiar to the type and energy of the incident particle and to the kind of interaction it undergoes. These partial cross-sections are summed to form the total cross-section; the ratio of the partial cross-section to the total cross-section gives the probability of this particular interaction occurring. Cross-section data for the interaction types of interest must be supplied for each material present. The model also consists of algorithms used to compute the result of interactions (changes in particle energy, direction, etc.) based on the physical principles that describe the interaction of radiation with matter and the cross-section data provided
Tissue-equivalent proportional counters (TEPC) can potentially be used as a portable and personal dosemeter in mixed neutron and gamma-ray fields, but what hinders this use is their typically large physical size. To formulate compact TEPC designs, the use of a Monte Carlo transport code is necessary to predict the performance of compact designs in these fields. To perform this modelling, three candidate codes were assessed: MCNPX 2.7.E, FLUKA 2011.2 and PHITS 2.24. In each code, benchmark simulations were performed involving the irradiation of a 5-in. TEPC with monoenergetic neutron fields and a 4-in. wall-less TEPC with monoenergetic gamma-ray fields. The frequency and dose mean lineal energies and dose distributions calculated from each code were compared with experimentally determined data. For the neutron benchmark simulations, PHITS produces data closest to the experimental values and for the gamma-ray benchmark simulations, FLUKA yields data closest to the experimentally determined quantities. (authors)
Development of a New Monte Carlo reactor physics code
Leppänen, Jaakko
2007-01-01
Monte Carlo neutron transport codes are widely used in various reactor physics applications, traditionally related to criticality safety analyses, radiation shielding problems, detector modelling and validation of deterministic transport codes. The main advantage of the method is the capability to model geometry and interaction physics without major approximations. The disadvantage is that the modelling of complicated systems is very computing-intensive, which restricts the applications to so...
SPQR: a Monte Carlo reactor kinetics code
The SPQR Monte Carlo code has been developed to analyze fast reactor core accident problems where conventional methods are considered inadequate. The code is based on the adiabatic approximation of the quasi-static method. This initial version contains no automatic material motion or feedback. An existing Monte Carlo code is used to calculate the shape functions and the integral quantities needed in the kinetics module. Several sample problems have been devised and analyzed. Due to the large statistical uncertainty associated with the calculation of reactivity in accident simulations, the results, especially at later times, differ greatly from deterministic methods. It was also found that in large uncoupled systems, the Monte Carlo method has difficulty in handling asymmetric perturbations
Coded aperture optimization using Monte Carlo simulations
Coded apertures using Uniformly Redundant Arrays (URA) have been unsuccessfully evaluated for two-dimensional and three-dimensional imaging in Nuclear Medicine. The images reconstructed from coded projections contain artifacts and suffer from poor spatial resolution in the longitudinal direction. We introduce a Maximum-Likelihood Expectation-Maximization (MLEM) algorithm for three-dimensional coded aperture imaging which uses a projection matrix calculated by Monte Carlo simulations. The aim of the algorithm is to reduce artifacts and improve the three-dimensional spatial resolution in the reconstructed images. Firstly, we present the validation of GATE (Geant4 Application for Emission Tomography) for Monte Carlo simulations of a coded mask installed on a clinical gamma camera. The coded mask modelling was validated by comparison between experimental and simulated data in terms of energy spectra, sensitivity and spatial resolution. In the second part of the study, we use the validated model to calculate the projection matrix with Monte Carlo simulations. A three-dimensional thyroid phantom study was performed to compare the performance of the three-dimensional MLEM reconstruction with conventional correlation method. The results indicate that the artifacts are reduced and three-dimensional spatial resolution is improved with the Monte Carlo-based MLEM reconstruction.
Monte Carlo solver for UWB1 nuclear fuel depletion code
Highlights: • A new Monte Carlo solver was developed in order to speed-up depletion calculations. • For LWR model, UWB1 Monte Carlo solver is on average 10 times faster than MCNP6. • The UWB1 code will allow faster calculation analysis of BA parameters in fuel design. - Abstract: Recent nuclear reactor burnable absorber research tries to introduce new materials in the nuclear fuel. As a part of this effort, a fast computational tool is being developed for the advanced nuclear fuel. The first version of the newly developed UWB1 fast nuclear fuel depletion code significantly reduced calculation time by omitting the solution step for the Boltzmann transport equation. However, estimation of neutron multiplication factor during depletion was not sufficiently calculated. Therefore, at least one transport calculation for fuel depletion is necessary. This paper presents a new Monte Carlo solver that is implemented into the UWB1 code. The UWB1 Monte Carlo solver calculates neutron multiplication factor and neutron flux in the fuel for collapsed cross sections. Accuracy of the solver is supported by using current nuclear data stored in the ENDF/B-VII.1 library. Speed of the solver is the product of development focusing on minimization of CPU utilization at the expense of RAM demands. The UWB1 Monte Carlo solver is approximately 14 times faster than the MCNP6 reference code when one transport equation solution within fuel depletion is compared. Another speed-up can be achieved by employing advanced depletion scheme in the coupled transport and burnup equations. The resulting faster code will be used in optimization studies for ideal burnable absorber material selection where many various materials and concentrations will be evaluated
Development of a New Monte Carlo reactor physics code
Monte Carlo neutron transport codes are widely used in various reactor physics applications, traditionally related to criticality safety analyses, radiation shielding problems, detector modelling and validation of deterministic transport codes. The main advantage of the method is the capability to model geometry and interaction physics without major approximations. The disadvantage is that the modelling of complicated systems is very computing-intensive, which restricts the applications to some extent. The importance of Monte Carlo calculation is likely to increase in the future, along with the development in computer capacities and parallel calculation. An interesting near-future application for the Monte Carlo method is the generation of input parameters for deterministic reactor simulator codes. These codes are used in coupled LWR full-core analyses and typically based on few-group nodal diffusion methods. The input data consists of homogenised few-group constants, presently generated using deterministic lattice transport codes. The task is becoming increasingly challenging, along with the development in nuclear technology. Calculations involving high-burnup fuels, advanced MOX technology and next-generation reactor systems are likely to cause problems in the future, if code development cannot keep up with the applications. A potential solution is the use of Monte Carlo based lattice transport codes, which brings all the advantages of the calculation method. So far there has been only a handful of studies on group constant generation using the Monte Carlo method, although the interest has clearly increased during the past few years. The homogenisation of reaction cross sections is simple and straightforward, and it can be carried out using any Monte Carlo code. Some of the parameters, however, require the use of special techniques that are usually not available in general-purpose codes. The main problem is the calculation of neutron diffusion coefficients, which
MCNP-POLIMI v1.0, Monte Carlo N-Particle Transport Code System To Simulate Time-Analysis Quantities
1 - Description of program or function: MCNP is a general-purpose, continuous-energy, generalized geometry, time-dependent, coupled neutron-photon-electron Monte Carlo transport code system. Based on the Los Alamos National Laboratory code MCNP4C (formerly distributed as CCC-700), MCNP-PoliMi was developed to simulate time-analysis quantities. In particular, the code includes the correlation between neutron interaction and the corresponding photon production. Conversely to the technique adopted by standard MCNP, MCNP PoliMi samples secondary photons according to the neutron collision type. A post-processing code, i.e. the Matlab script 'postmain', is included and can be tailored to model specific detector characteristics. These features make MCNP-PoliMi a versatile tool to simulate particle interactions and detection processes. 2 - Methods: MCNP treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces. For neutrons, all reactions in a particular cross-section evaluation are accounted for. Both free gas and S(alpha, beta) thermal treatments are used. Criticality sources as well as fixed and surface sources are available. For photons, the code takes account of incoherent and coherent scattering with and without electron binding effects, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. A very general source and tally structure is available. The tallies have extensive statistical analysis of convergence. Rapid convergence is enabled by a wide variety of variance reduction methods. Energy ranges are 0-60 MeV for neutrons (data generally only available up to 20 MeV) and 1 keV - 1 GeV for photons and electrons. The MCNP-PoliMi code was developed to simulate each neutron-nucleus interaction as closely as possible. In particular, neutron interaction and
Objective: To discuss the feasibility of Monte Carlo N-particle transport code (MCNP) simulated calculation. Methods: The calculation in water phantom was contrasted with the practical measurements and the reported values using the percent depth dose (PDD) curve and normal peak scatter factor. Results: There Was no significant difference between calculated and measured results in the 10 cm×10 cm field (t=-0.41, P>0.05), however, there were significant differences in the 5 cm×5 cm field (t=7.2, P<0.05) and in the 12 cm×12 cm field (t=-4.6, P<0.05). There was no significant difference between the calculated results and the reported values (t=-1.91, P>0.05). In the same radiation field, the PDD decreased as the depth increased, but increased as the size of the radiation field increased at the same depth. PDD and normal peak scatter factor were both important parameters for calculation of prescribed dose. Conclusions: It is possible to establish a set of accurate and comprehensive percent depth doses and normal peak scatter factor parameters so as to provide the basis of clinical use, quality assurance and quality control for radiotherapy. (authors)
Implict Monte Carlo Radiation Transport Simulations of Four Test Problems
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.
The crucial problem for radiation shielding design at heavy-ion accelerator facilities with beam energies to several GeV/n is the source term problem. Experimental data on double differential neutron yields from thick target irradiated with high-energy uranium nuclei are lacking. At present, there are not many Monte-Carlo multipurpose codes that can work with primary high-energy uranium nuclei. These codes use different physical models for simulation of nucleus-nucleus reactions. Therefore, verification of the codes with available experimental data is very important for selection of the most reliable code for practical tasks. This paper presents comparisons of the FLUKA, GEANT4 and SHIELD codes simulations with the experimental data on neutron production at 1 GeV/n 238U beam interaction with thick Fe target
Highlights: • We investigated the main quantities determining ADS performance. • We calculated ADS performance such as neutron yield, neutron leakage spectra, heating and neutron and proton spectra in the target and in the beam window. • We used the MCNPX-2.7.0 Monte Carlo code for three-dimensional calculations. - Abstract: The MCNPX code offers options based on physics packages; the Bertini, ISABEL, INCL4 intra-nuclear models, and Dresner, ABLA evaporation–fission models and CEM2k cascade-exciton model. This study analyzes the main quantities determining ADS performance, such as neutron yield, neutron leakage spectra, heating and neutron and proton spectra in the target and in the beam window calculated by the MCNPX-2.5.0 Monte Carlo transport code, which is a combination of LAHET and MCNP codes. The results obtained by simulating different models cited above and implemented in MCNPX are compared with each other. The investigated system is composed of a natural lead cylindrical target and stainless steel (HT9) beam window. The target has been optimized to produce maximum number of neutrons with a radius of 20 cm and 70 cm of height. The target is bombarded with a high intensity linear accelerator by a 1 GeV, 1 mA proton beam. The protons are assumed uniformly distributed across the beam of radius 3 cm, and entering the target through a hole of 5.3 cm radius. The proton beam has an outer radius of 5.3 cm and an inner radius of 5.0 cm. The maximum value of the neutron flux in the target is observed on the axis ∼10 cm below the beam window, where the maximum difference between 7 different models is ∼15%. The total neutron leakage of the target calculated with the Bertini/ABLA is 1.83 × 1017 n/s, and is about 14% higher than the value calculated by the INCL4/Dresner (1.60 × 1017 n/s). Bertini/ABLA calculates top, bottom and side neutron leakage fractions as 20%, 2.3%, 77.6% of the total leakage, respectively, whereas, the calculated fractions are 18
Development of Monte Carlo machine for particle transport problem
Monte Carlo machine, Monte-4 has been developed to realize high performance computing of Monte Carlo codes for particle transport. The calculation for particle tracking in a complex geometry requires (1) classification of particles by the region types using multi-way conditional branches, and (2) determination whether intersections of particle paths with surfaces of the regions are on the boundaries of the regions or not, using nests of conditional branches. However, these procedures require scalar operations or unusual vector operations. Thus the speedup ratios have been low, i.e. nearly two times, in vector processing of Monte Carlo codes for particle transport on conventional vector processors. The Monte Carlo machine Monte-4 has been equipped with the special hardware called Monte Carlo pipelines to process these procedures with high performance. Additionally Monte-4 has been equipped with enhanced load/store pipelines to realize fast transfer of indirectly addressed data for the purpose of resolving imbalances between the performance of data transfers and arithmetic operations in vector processing of Monte Carlo codes on conventional vector processors. Finally, Monte-4 has a parallel processing capability with four processors to multiply the performance of vector processing. We have evaluated the effective performance of Monte-4 using production-level Monte Carlo codes such as vectorized KENO-IV and MCNP. In the performance evaluation, nearly ten times speedup ratios have been obtained, compared with scalar processing of the original codes. (author)
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 ^{12}C 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.
The simulations for linear devices using the 3D Monte-Carlo code ERO have a number of particularities of both physical and technical nature. Many physical effects introduced and tested for these simulations (elastic collisions, population of metastable states, etc.) can be relevant for tokamaks as well. The current status of these activities is summarized. As an example some simulations reproducing the experimental observations at PISCES-B and Pilot-PSI are presented. Main deviations and ideas for their origins are discussed (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)
Criticality benchmarking of ANET Monte Carlo code
In this work the new Monte Carlo code ANET is tested on criticality calculations. ANET is developed based on the high energy physics code GEANT of CERN and aims at progressively satisfying several requirements regarding both simulations of GEN II/III reactors, as well as of innovative nuclear reactor designs such as the Accelerator Driven Systems (ADSs). Here ANET is applied on three different nuclear configurations, including a subcritical assembly, a Material Testing Reactor and the conceptual configuration of an ADS. In the first case, calculation of the effective multiplication factor (keff) are performed for the Training Nuclear Reactor of the Aristotle University of Thessaloniki, while in the second case keff is computed for the fresh fueled core of the Portuguese research reactor (RPJ) just after its conversion to Low Enriched Uranium, considering the control rods at the position that renders the reactor critical. In both cases ANET computations are compared with corresponding results obtained by three different well established codes, including both deterministic (XSDRNPM/CITATION) and Monte Carlo (TRIPOLI, MCNP). In the RPI case, keff computations are also compared with observations during the reactor core commissioning since the control rods are considered at criticality position. The above verification studies show ANET to produce reasonable results since they are satisfactorily compared with other models as well as with observations. For the third case (ADS), preliminary ANET computations of keff for various intensities of the proton beam are presented, showing also a reasonable code performance concerning both the order of magnitude and the relative variation of the computed parameter. (author)
Vilches, M.; García-Pareja, S.; Guerrero, R.; Anguiano, M.; Lallena, A. M.
2007-09-01
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.
Proton therapy Monte Carlo SRNA-VOX code
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.
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.
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 opti...
Monte Carlo methods for particle transport
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 ...
Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method
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.
Status of the MCNPX transport code
The Monte Carlo particle transport code MCNPX and its associated data have been the focus of a major development effort at Los Alamos for several years. The system has reached a mature state, and has become a significant tool for many intermediate and high-energy particle transport applications. A recent version has been released to the Radiation Safety Information Computational Center (RSICC). A recent report provides an overview of the code and an extensive set of references for the component physics modules used in the code. In this paper we review the status of the developmental version of MCNPX, and describe some important new enhancements, including the use of evaluated nuclear data files for proton transport; the use of photonuclear reaction data; improved elastic and inelastic react ion cross sections for nucleons, antinucleons, pions, and kaons; and two new modes of operation of the code. We also illustrate the use of the new proton and photonuclear data in two representative applications
Recent developments in the Los Alamos radiation transport code system
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.
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.)
Parallel MCNP Monte Carlo transport calculations with MPI
The steady increase in computational performance has made Monte Carlo calculations for large/complex systems possible. However, in order to make these calculations practical, order of magnitude increases in performance are necessary. The Monte Carlo method is inherently parallel (particles are simulated independently) and thus has the potential for near-linear speedup with respect to the number of processors. Further, the ever-increasing accessibility of parallel computers, such as workstation clusters, facilitates the practical use of parallel Monte Carlo. Recognizing the nature of the Monte Carlo method and the trends in available computing, the code developers at Los Alamos National Laboratory implemented the message-passing general-purpose Monte Carlo radiation transport code MCNP (version 4A). The PVM package was chosen by the MCNP code developers because it supports a variety of communication networks, several UNIX platforms, and heterogeneous computer systems. This PVM version of MCNP has been shown to produce speedups that approach the number of processors and thus, is a very useful tool for transport analysis. Due to software incompatibilities on the local IBM SP2, PVM has not been available, and thus it is not possible to take advantage of this useful tool. Hence, it became necessary to implement an alternative message-passing library package into MCNP. Because the message-passing interface (MPI) is supported on the local system, takes advantage of the high-speed communication switches in the SP2, and is considered to be the emerging standard, it was selected
High Energy Transport Code HETC
The physics contained in the High Energy Transport Code (HETC), in particular the collision models, are discussed. An application using HETC as part of the CALOR code system is also given. 19 refs., 5 figs., 3 tabs
Weighted-delta-tracking for Monte Carlo particle transport
Highlights: • This paper presents an alteration to the Monte Carlo Woodcock tracking technique. • The alteration improves computational efficiency within regions of high absorbers. • The rejection technique is replaced by a statistical weighting mechanism. • The modified Woodcock method is shown to be faster than standard Woodcock tracking. • The modified Woodcock method achieves a lower variance, given a specified accuracy. - Abstract: Monte Carlo particle transport (MCPT) codes are incredibly powerful and versatile tools to simulate particle behavior in a multitude of scenarios, such as core/criticality studies, radiation protection, shielding, medicine and fusion research to name just a small subset applications. However, MCPT codes can be very computationally expensive to run when the model geometry contains large attenuation depths and/or contains many components. This paper proposes a simple modification to the Woodcock tracking method used by some Monte Carlo particle transport codes. The Woodcock method utilizes the rejection method for sampling virtual collisions as a method to remove collision distance sampling at material boundaries. However, it suffers from poor computational efficiency when the sample acceptance rate is low. The proposed method removes rejection sampling from the Woodcock method in favor of a statistical weighting scheme, which improves the computational efficiency of a Monte Carlo particle tracking code. It is shown that the modified Woodcock method is less computationally expensive than standard ray-tracing and rejection-based Woodcock tracking methods and achieves a lower variance, given a specified accuracy
The present of shielding analysis with nuclear data for continuous energy Monte Carlo code MCNP
Following three problems are analyzed by continuous energy Monte Carlo code MCNP with JENDL-3.2, 3.3, and ENDF/B-VI. 1. Shielding analysis of WINFRITH-Aspins iron deep penetration experiment. 2. Shielding analysis of TN-12A spent fuel transport cask experiment. 3. Shielding analysis of modular shielding house keeping spent fuel transportable casks. (author)
MCNP4B-GN, Monte Carlo Code System for (gamma,n) production and transport in high-Z materials
1 - Description of program or function: MCNP4B-GN is used to treat (gamma,n) production and transport in medical accelerator heads, to study the undesired neutron dose to patients, employing a single code for both the electromagnetic and the neutron transport. 2 - Methods: The code simulates the production of giant dipole resonance (GDR) photoneutrons in thick layers of high-Z elements. Neutrons are generated through evaporation of the compound nucleus or through direct channel; the photoneutron origin coordinates, evaluated as the electromagnetic shower develops are taken into account as well as the energy spectrum of the generated photoneutron. Photoneutron production routines have been inserted into MCNP4B, thus allowing a handling of complicated geometries with a single input definition, a fundamental requirement for this kind of application. The modifications to the standard MCNP4B were made as a 'patch' (i.e. a series of instructions on how to modify the basic code) which is distributed together with a preprocessor. The preprocessor reads the patch, reads MCNP4B and writes MCNP4B-GN. A new subroutine 'GAMMN' has been written, analogous to the subroutine 'ACEGAM' for (n,gamma). GAMMN is called from subroutine 'COLIDP' at a photon collision with one of the 6 elements in question when the energy of the photon is such that there is a non-zero probability of producing a neutron. In analogy with the (n,gamma) capability, the production or not of a neutron is not correlated with the subsequent history of the photon and in particular the choice of the type of photon collision at that spatial point. 3 - Restrictions on the complexity of the problem: The upper energy limit, imposed by theoretical considerations, is 30 MeV for photons and 20 MeV for photoneutrons. The physical model breaks down at higher photon energies, whilst the neutron transport performed with MCNP cannot be simulated for neutron energies greater than 20 MeV (the neutron energy regime in MCNP is
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.
Development of Monte Carlo-based pebble bed reactor fuel management code
Highlights: • A new Monte Carlo-based fuel management code for OTTO cycle pebble bed reactor was developed. • The double-heterogeneity was modeled using statistical method in MVP-BURN code. • The code can perform analysis of equilibrium and non-equilibrium phase. • Code-to-code comparisons for Once-Through-Then-Out case were investigated. • Ability of the code to accommodate the void cavity was confirmed. - Abstract: A fuel management code for pebble bed reactors (PBRs) based on the Monte Carlo method has been developed in this study. The code, named Monte Carlo burnup analysis code for PBR (MCPBR), enables a simulation of the Once-Through-Then-Out (OTTO) cycle of a PBR from the running-in phase to the equilibrium condition. In MCPBR, a burnup calculation based on a continuous-energy Monte Carlo code, MVP-BURN, is coupled with an additional utility code to be able to simulate the OTTO cycle of PBR. MCPBR has several advantages in modeling PBRs, namely its Monte Carlo neutron transport modeling, its capability of explicitly modeling the double heterogeneity of the PBR core, and its ability to model different axial fuel speeds in the PBR core. Analysis at the equilibrium condition of the simplified PBR was used as the validation test of MCPBR. The calculation results of the code were compared with the results of diffusion-based fuel management PBR codes, namely the VSOP and PEBBED codes. Using JENDL-4.0 nuclide library, MCPBR gave a 4.15% and 3.32% lower keff value compared to VSOP and PEBBED, respectively. While using JENDL-3.3, MCPBR gave a 2.22% and 3.11% higher keff value compared to VSOP and PEBBED, respectively. The ability of MCPBR to analyze neutron transport in the top void of the PBR core and its effects was also confirmed
MCMG: a 3-D multigroup P3 Monte Carlo code and its benchmarks
In this paper a 3-D Monte Carlo multigroup neutron transport code MCMG has been developed from a coupled neutron and photon transport Monte Carlo code MCNP. The continuous-energy cross section library of the MCNP code is replaced by the multigroup cross section data generated by the transport lattice code, such as the WIMS code. It maintains the strong abilities of MCNP for geometry treatment, counting, variance reduction techniques and plotting. The multigroup neutron scattering cross sections adopt the Pn (n ≤ 3) approximation. The test results are in good agreement with the results of other methods and experiments. The number of energy groups can be varied from few groups to multigroup, and either macroscopic or microscopic cross section can be used. (author)
ITS: integrated TIGER series of electron/photon transport codes
The ITS system of linear, time-independent coupled electron/photon Monte Carlo transport codes is a group of multimaterial, multidimensional codes designed to provide an accurate description of the production and transport of the electron/photon cascade, with or without the presence of macroscopic electric and magnetic fields. It is based primarily on the ETRAN system, which combines microscopic photon transport with a macroscopic random walk for electron transport. Emphasis is on simplicity of application without sacrificing sophistication of the physical model. Flexibility of construction permits the user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications
Scalable Domain Decomposed Monte Carlo Particle Transport
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.
Fast code for Monte Carlo simulations
A computer code to generate the dynamic evolution of the Ising model on a square lattice, following the Metropolis algorithm is presented. The computer time consumption is reduced by a factor of 8 when one compares our code with traditional multiple spin codes. The memory allocation size is also reduced by a factor of 4. The code is easily generalizable for other lattices and models. (author)
In the framework of French research program on Generation IV sodium cooled fast reactor, one possible option consists in burning minor actinides in this kind of Advanced Sodium Technological Reactor. Two types of transmutation mode are studied in the world : the homogeneous mode of transmutation where actinides are scattered with very low enrichment ratio in fissile assemblies and the heterogeneous mode where fissile core is surrounded by blanket assemblies filled with minor actinides with ratio of incorporated actinides up to 20%. Depending on which element is considered to be burnt and on its content, these minor actinides contents imply constraints on assemblies' transportation between Nuclear Power Plants and fuel cycle facilities. In this study, we present some academic studies in order to identify some key constraints linked to the residual power and neutron/gamma load of such kind of blanket assemblies. To simplify the approach, we considered a modeling of a 'model cask' dedicated to the transportation of a unique irradiated blanket assembly loaded with 20% of Americium and basically inspired from an existent cask designed initially for the damaged fissile Superphenix assembly transport. Thermal calculations performed with EDF-SYRTHES code have shown that due to thermal limitations on cladding temperature, the decay time to be considered before transportation is 20 years. This study is based on explicit 3D representations of the cask and the contained blanket assembly with the Monte Carlo code TRIPOLI/JEFF3.1.1 library and concludes that after such a decay time, the transportation of a unique Americium radial blanket is feasible only if the design of our model cask is modified in order to comply with the dose limitation criterion. (author)
Applications of the Monte Carlo radiation transport toolkit at LLNL
Sale, Kenneth E.; Bergstrom, Paul M., Jr.; Buck, Richard M.; Cullen, Dermot; Fujino, D.; Hartmann-Siantar, Christine
1999-09-01
Modern Monte Carlo radiation transport codes can be applied to model most applications of radiation, from optical to TeV photons, from thermal neutrons to heavy ions. Simulations can include any desired level of detail in three-dimensional geometries using the right level of detail in the reaction physics. The technology areas to which we have applied these codes include medical applications, defense, safety and security programs, nuclear safeguards and industrial and research system design and control. The main reason such applications are interesting is that by using these tools substantial savings of time and effort (i.e. money) can be realized. In addition it is possible to separate out and investigate computationally effects which can not be isolated and studied in experiments. In model calculations, just as in real life, one must take care in order to get the correct answer to the right question. Advancing computing technology allows extensions of Monte Carlo applications in two directions. First, as computers become more powerful more problems can be accurately modeled. Second, as computing power becomes cheaper Monte Carlo methods become accessible more widely. An overview of the set of Monte Carlo radiation transport tools in use a LLNL will be presented along with a few examples of applications and future directions.
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...... in throughput that exceed 10x for TCP traffic while requiring a limited sacrifice in terms of fairness towards other flows on the channel....
Monte Carlo simulation of medical linear accelerator using primo code
The use of monte Carlo simulation has become very important in the medical field and especially in calculation in radiotherapy. Various Monte Carlo codes were developed simulating interactions of particles and photons with matter. One of these codes is PRIMO that performs simulation of radiation transport from the primary electron source of a linac to estimate the absorbed dose in a water phantom or computerized tomography (CT). PRIMO is based on Penelope Monte Carlo code. Measurements of 6 MV photon beam PDD and profile were done for Elekta precise linear accelerator at Radiation and Isotopes Center Khartoum using computerized Blue water phantom and CC13 Ionization Chamber. accept Software was used to control the phantom to measure and verify dose distribution. Elektalinac from the list of available linacs in PRIMO was tuned to model Elekta precise linear accelerator. Beam parameter of 6.0 MeV initial electron energy, 0.20 MeV FWHM, and 0.20 cm focal spot FWHM were used, and an error of 4% between calculated and measured curves was found. The buildup region Z max was 1.40 cm and homogenous profile in cross line and in line were acquired. A number of studies were done to verily the model usability one of them is the effect of the number of histories on accuracy of the simulation and the resulted profile for the same beam parameters. The effect was noticeable and inaccuracies in the profile were reduced by increasing the number of histories. Another study was the effect of Side-step errors on the calculated dose which was compared with the measured dose for the same setting.It was in range of 2% for 5 cm shift, but it was higher in the calculated dose because of the small difference between the tuned model and measured dose curves. Future developments include simulating asymmetrical fields, calculating the dose distribution in computerized tomographic (CT) volume, studying the effect of beam modifiers on beam profile for both electron and photon beams.(Author)
MORET: Version 4.B. A multigroup Monte Carlo criticality code
MORET 4 is a three dimensional multigroup Monte Carlo code which calculates the effective multiplication factor (keff) of any configurations more or less complex as well as reaction rates in the different volumes of the geometry and the leakage out of the system. MORET 4 is the Monte Carlo code of the APOLLO2-MORET 4 standard route of CRISTAL, the French criticality package. It is the most commonly used Monte Carlo code for French criticality calculations. During the last four years, the MORET 4 team has developed or improved the following major points: modernization of the geometry, implementation of perturbation algorithms, source distribution convergence, statistical detection of stationarity, unbiased variance estimation and creation of pre-processing and post-processing tools. The purpose of this paper is not only to present the new features of MORET but also to detail clearly the physical models and the mathematical methods used in the code. (author)
MOx benchmark calculations by deterministic and Monte Carlo codes
Highlights: ► MOx based depletion calculation. ► Methodology to create continuous energy pseudo cross section for lump of minor fission products. ► Mass inventory comparison between deterministic and Monte Carlo codes. ► Higher deviation was found for several isotopes. - Abstract: A depletion calculation benchmark devoted to MOx fuel is an ongoing objective of the OECD/NEA WPRS following the study of depletion calculation concerning UOx fuels. The objective of the proposed benchmark is to compare existing depletion calculations obtained with various codes and data libraries applied to fuel and back-end cycle configurations. In the present work the deterministic code NEWT/ORIGEN-S of the SCALE6 codes package and the Monte Carlo based code MONTEBURNS2.0 were used to calculate the masses of inventory isotopes. The methodology to apply the MONTEBURNS2.0 to this benchmark is also presented. Then the results from both code were compared.
Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations
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.
Recent transportation calculation code systems and their accuracy evaluation
In the field of shielding design, many studies have been carried out for the development of radiation transportation calculation codes (transportation codes) including Monte Carlo codes. The present report outlines major transportation codes used in Japan for design of shielding. Major one-dimensional codes include ANISN (Sn), PALLAS-PL and SP-Br (direct integration) whili two-dimensional ones include DOT-3.5 and TWOTRAN-II. All these transportation codes have been developed on the basis of numerical solution to the Boltzmann's transportation equation. These codes are roughly divided into two groups: discrete ordinates type and Monte Carlo type. The former include Sn-type codes and direct integration type codes. Sn-type codes are currently used most widely. The accuracy and other features of a code should be tested before applysing it to practical shielding design. One of the techniques for this purpose is the benchmark method, which consists of benchmark tests and analysis of the test results. The possible overall error involved in calculations can be determined from the benchmark tests. (Nogami, K.)
A Monte Carlo track structure code for low energy protons
Endo, S; Nikjoo, H; Uehara, S; Hoshi, M; Ishikawa, M; Shizuma, K
2002-01-01
A code is described for simulation of protons (100 eV to 10 MeV) track structure in water vapor. The code simulates molecular interaction by interaction for the transport of primary ions and secondary electrons in the form of ionizations and excitations. When a low velocity ion collides with the atoms or molecules of a target, the ion may also capture or lose electrons. The probabilities for these processes are described by the quantity cross-section. Although proton track simulation at energies above Bragg peak (>0.3 MeV) has been achieved to a high degree of precision, simulations at energies near or below the Bragg peak have only been attempted recently because of the lack of relevant cross-section data. As the hydrogen atom has a different ionization cross-section from that of a proton, charge exchange processes need to be considered in order to calculate stopping power for low energy protons. In this paper, we have used state-of-the-art Monte Carlo track simulation techniques, in conjunction with the pub...
The Monte Carlo code MCSHAPE: Main features and recent developments
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
Benchmarking Monte Carlo codes for criticality safety using subcritical measurements
Monte Carlo codes that are used for criticality safety evaluations are typically validated using critical experiments in which the neutron multiplication factor is unity. However, the conditions for most fissile material operations do not coincide to those of the critical experiments. This paper demonstrates that Monte Carlo methods and nuclear data can be validated using subcritical measurements whose conditions may coincide more closely to actual configurations of fissile material. (orig.)
Design of shielding of LILW containers by Monte Carlo codes
Accurate prediction of dose rates from containers with radioactive waste is becoming more important regarding more rigorous regulative in this area. The usual approach to the problem consists in combining numerical and measuring methods. In this paper a Monte Carlo calculations were used for calculating doses from a standard 200 liter drum which contains the intermediate level radioactive waste. Two different Monte Carlo codes were applied and compared, for the same combination of parameters. (author)
Development of 3d reactor burnup code based on Monte Carlo method and exponential Euler method
Burnup analysis plays a key role in fuel breeding, transmutation and post-processing in nuclear reactor. Burnup codes based on one-dimensional and two-dimensional transport method have difficulties in meeting the accuracy requirements. A three-dimensional burnup analysis code based on Monte Carlo method and Exponential Euler method has been developed. The coupling code combines advantage of Monte Carlo method in complex geometry neutron transport calculation and FISPACT in fast and precise inventory calculation, meanwhile resonance Self-shielding effect in inventory calculation can also be considered. The IAEA benchmark text problem has been adopted for code validation. Good agreements were shown in the comparison with other participants' results. (authors)
JCOGIN. A parallel programming infrastructure for Monte Carlo particle transport
The advantages of the Monte Carlo method for reactor analysis are well known, but the full-core reactor analysis challenges the computational time and computer memory. Meanwhile, the exponential growth of computer power in the last 10 years is now creating a great opportunity for large scale parallel computing on the Monte Carlo full-core reactor analysis. In this paper, a parallel programming infrastructure is introduced for Monte Carlo particle transport, named JCOGIN, which aims at accelerating the development of Monte Carlo codes for the large scale parallelism simulations of the full-core reactor. Now, JCOGIN implements the hybrid parallelism of the spatial decomposition and the traditional particle parallelism on MPI and OpenMP. Finally, JMCT code is developed on JCOGIN, which reaches the parallel efficiency of 70% on 20480 cores for fixed source problem. By the hybrid parallelism, the full-core pin-by-pin simulation of the Dayawan reactor was implemented, with the number of the cells up to 10 million and the tallies of the fluxes utilizing over 40GB of memory. (author)
Hybrid Deterministic-Monte Carlo Methods for Neutral Particle Transport
In the history of transport analysis methodology for nuclear systems, there have been two fundamentally different methods, i.e., deterministic and Monte Carlo (MC) methods. Even though these two methods coexisted for the past 60 years and are complementary each other, they never been coded in the same computer codes. Recently, however, researchers have started to consider to combine these two methods in a computer code to make use of the strengths of two algorithms and avoid weaknesses. Although the advanced modern deterministic techniques such as method of characteristics (MOC) can solve a multigroup transport equation very accurately, there are still uncertainties in the MOC solutions due to the inaccuracy of the multigroup cross section data caused by approximations in the process of multigroup cross section generation, i.e., equivalence theory, interference effects, etc. Conversely, the MC method can handle the resonance shielding effect accurately when sufficiently many neutron histories are used but it takes a long calculation time. There was also a research to combine a multigroup transport and a continuous energy transport solver in a computer code system depending on the energy range. This paper proposes a hybrid deterministic-MC method in which a multigroup MOC method is used for high and low energy range and continuous MC method is used for the intermediate resonance energy range for efficient and accurate transport analysis
MCOR - Monte Carlo depletion code for reference LWR calculations
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
MCOR - Monte Carlo depletion code for reference LWR calculations
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
Monte Carlo method in radiation transport problems
In neutral radiation transport problems (neutrons, photons), two values are important: the flux in the phase space and the density of particles. To solve the problem with Monte Carlo method leads to, among other things, build a statistical process (called the play) and to provide a numerical value to a variable x (this attribution is called score). Sampling techniques are presented. Play biasing necessity is proved. A biased simulation is made. At last, the current developments (rewriting of programs for instance) are presented due to several reasons: two of them are the vectorial calculation apparition and the photon and neutron transport in vacancy media
Adjoint electron-photon transport Monte Carlo calculations with ITS
A general adjoint coupled electron-photon Monte Carlo code for solving the Boltzmann-Fokker-Planck equation has recently been created. It is a modified version of ITS 3.0, a coupled electronphoton Monte Carlo code that has world-wide distribution. The applicability of the new code to radiation-interaction problems of the type found in space environments is demonstrated
Verification of the shift Monte Carlo code with the C5G7 reactor benchmark
Sly, N. C.; Mervin, B. T. [Dept. of Nuclear Engineering, Univ. of Tennessee, 311 Pasqua Engineering Building, Knoxville, TN 37996-2300 (United States); Mosher, S. W.; Evans, T. M.; Wagner, J. C. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, 311 Pasqua Engineering Building, Knoxville, TN 37996-2300 (United States)
2012-07-01
Shift is a new hybrid Monte Carlo/deterministic radiation transport code being developed at Oak Ridge National Laboratory. At its current stage of development, Shift includes a parallel Monte Carlo capability for simulating eigenvalue and fixed-source multigroup transport problems. This paper focuses on recent efforts to verify Shift's Monte Carlo component using the two-dimensional and three-dimensional C5G7 NEA benchmark problems. Comparisons were made between the benchmark eigenvalues and those output by the Shift code. In addition, mesh-based scalar flux tally results generated by Shift were compared to those obtained using MCNP5 on an identical model and tally grid. The Shift-generated eigenvalues were within three standard deviations of the benchmark and MCNP5-1.60 values in all cases. The flux tallies generated by Shift were found to be in very good agreement with those from MCNP. (authors)
Dehaye, Benjamin,
2014-01-01
Fields such as criticality studies need to compute some values of interest in neutron physics. Two kind of codes may be used : deterministic ones and stochastic ones. The stochastic codes do not require approximation and are thus more exact. However, they may require a lot of time to converge with a sufficient precision.The work carried out during this thesis aims to build an efficient acceleration strategy in the TRIPOLI-4®. We wish to implement the zero variance game. To do so, the method r...
PEREGRINE: An all-particle Monte Carlo code for radiation therapy
The goal of radiation therapy is to deliver a lethal dose to the tumor while minimizing the dose to normal tissues. To carry out this task, it is critical to calculate correctly the distribution of dose delivered. Monte Carlo transport methods have the potential to provide more accurate prediction of dose distributions than currently-used methods. PEREGRINE is a new Monte Carlo transport code developed at Lawrence Livermore National Laboratory for the specific purpose of modeling the effects of radiation therapy. PEREGRINE transports neutrons, photons, electrons, positrons, and heavy charged-particles, including protons, deuterons, tritons, helium-3, and alpha particles. This paper describes the PEREGRINE transport code and some preliminary results for clinically relevant materials and radiation sources
Differential measurements and transport codes
A lot of projects involving particle beams of intermediate energy are calculated by using computerized simulation. TIERCE code system is dedicated to particle transport and is made up of 2 main codes HETC and MCNP, linked to their specialized nuclear data libraries. Here we present some results from TIERCE and we compare them with measured values. The quality of such simulation greatly depends on the nuclear reaction models used in the codes, so an important effort is made to improve these models. (A.C.)
Proton therapy Monte Carlo SRNA-VOX code
Ilić Radovan D.
2012-01-01
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 Monte Carlo code for ion beam therapy
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...
Computer codes in particle transport physics
Simulation of transport and interaction of various particles in complex media and wide energy range (from 1 MeV up to 1 TeV) is very complicated problem that requires valid model of a real process in nature and appropriate solving tool - computer code and data library. A brief overview of computer codes based on Monte Carlo techniques for simulation of transport and interaction of hadrons and ions in wide energy range in three dimensional (3D) geometry is shown. Firstly, a short attention is paid to underline the approach to the solution of the problem - process in nature - by selection of the appropriate 3D model and corresponding tools - computer codes and cross sections data libraries. Process of data collection and evaluation from experimental measurements and theoretical approach to establishing reliable libraries of evaluated cross sections data is Ion g, difficult and not straightforward activity. For this reason, world reference data centers and specialized ones are acknowledged, together with the currently available, state of art evaluated nuclear data libraries, as the ENDF/B-VI, JEF, JENDL, CENDL, BROND, etc. Codes for experimental and theoretical data evaluations (e.g., SAMMY and GNASH) together with the codes for data processing (e.g., NJOY, PREPRO and GRUCON) are briefly described. Examples of data evaluation and data processing to generate computer usable data libraries are shown. Among numerous and various computer codes developed in transport physics of particles, the most general ones are described only: MCNPX, FLUKA and SHIELD. A short overview of basic application of these codes, physical models implemented with their limitations, energy ranges of particles and types of interactions, is given. General information about the codes covers also programming language, operation system, calculation speed and the code availability. An example of increasing computation speed of running MCNPX code using a MPI cluster compared to the code sequential option
AGENT code - neutron transport benchmark examples
The paper focuses on description of representative benchmark problems to demonstrate the versatility and accuracy of the AGENT (Arbitrary Geometry Neutron Transport) code. AGENT couples the method of characteristics and R-functions allowing true modeling of complex geometries. AGENT is optimized for robustness, accuracy, and computational efficiency for 2-D assembly configurations. The robustness of R-function based geometry generator is achieved through the hierarchical union of the simple primitives into more complex shapes. The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through true geometries. The computational efficiency is maintained through a set of acceleration techniques introduced in all important calculation levels. The selected assembly benchmark problems discussed in this paper are: the complex hexagonal modular high-temperature gas-cooled reactor, the Purdue University reactor and the well known C5G7 benchmark model. (author)
The Monte Carlo code TRAMO - Capabilities and instructions for application
The report is intended for readers familiar with the fundamentals of the Monte Carlo method. Those readers might be interested in learning about successful generalisations as well as new ideas for curbing the statistical errors involved. Another intention however is to explain the significant basic features of the multigroup Monte Carlo code TRAMO, including the required input, so that readers will be able to performing the required adjustments to the specific calculation technique and develop their own tools for performing their specific calculations. An indispensable code needed for such TRAMO applications is the TRAWEI Monte Carlo code which calculates he required weightings for applications of the variance reducing Weight Window Method; other codes required are those for generating the neutron cross-section data and the group data. The TRAMO code calculates, with given source distribution of neutrons in multigroup approximation, multigroup flux data, integrated group flux data, and dose values for given partial volumes and surfaces. There are further code versions for calculation of neutron and gamma fluxes, or criticality data, but these are not considered in the report. (orig./CB)
Taylor series development in the Monte Carlo code Tripoli-4
Mazzolo, Alain; Zoia, Andrea; Martin, Brunella
2014-06-01
Perturbation methods for one or several variables based on the Taylor series development up to the second order is presented for the collision estimator in the framework of the Monte Carlo code Tripoli-4. Comparisons with the correlated sampling method implemented in Tripoli-4 demonstrate the need of including the cross derivatives in the development.
Discrete angle biasing in Monte Carlo radiation transport
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//0 used in the combined biasing schemes
Monte Carlo simulations of neoclassical transport in toroidal plasmas
FORTEC-3D code, which solves the drift-kinetic equation for torus plasmas and radial electric field using the δf Monte Carlo method, has developed to study the variety of issues relating to neoclassical transport phenomena in magnetic confinement plasmas. Here the numerical techniques used in FORTEC-3D are reviewed, and resent progress in the simulation method to simulate GAM oscillation is also explained. A band-limited white noise term is introduced in the equation of time evolution of radial electric field to excite GAM oscillation, which enables us to analyze GAM frequency using FORTEC-3D even in the case the collisionless GAM damping is fast. (author)
Monte Carlo methods in electron transport problems. Pt. 1
The condensed-history Monte Carlo method for charged particles transport is reviewed and discussed starting from a general form of the Boltzmann equation (Part I). The physics of the electronic interactions, together with some pedagogic example will be introduced in the part II. The lecture is directed to potential users of the method, for which it can be a useful introduction to the subject matter, and wants to establish the basis of the work on the computer code RECORD, which is at present in a developing stage
ALEPH 1.1.2: A Monte Carlo burn-up code
In the last 40 years, Monte Carlo particle transport has been applied to a multitude of problems such as shielding and medical applications, to various types of nuclear reactors, . . . The success of the Monte Carlo method is mainly based on its broad application area, on its ability to handle nuclear data not only in its most basic but also most complex form (namely continuous energy cross sections, complex interaction laws, detailed energy-angle correlations, multi-particle physics, . . . ), on its capability of modeling geometries from simple 1D to complex 3D, . . . There is also a current trend in Monte Carlo applications toward high detail 3D calculations (for instance voxel-based medical applications), something for which deterministic codes are neither suited nor performant as to computational time and precision. Apart from all these fields where Monte Carlo particle transport has been applied successfully, there is at least one area where Monte Carlo has had limited success, namely burn-up and activation calculations where the time parameter is added to the problem. The concept of Monte Carlo burn-up consists of coupling a Monte Carlo code to a burn-up module to improve the accuracy of depletion and activation calculations. For every time step the Monte Carlo code will provide reaction rates to the burn-up module which will return new material compositions to the Monte Carlo code. So if static Monte Carlo particle transport is slow, then Monte Carlo particle transport with burn-up will be even slower as calculations have to be performed for every time step in the problem. The computational issues to perform accurate Monte Carlo calculations are however continuously reduced due to improvements made in the basic Monte Carlo algorithms, due to the development of variance reduction techniques and due to developments in computer architecture (more powerful processors, the so-called brute force approach through parallel processors and networked systems
CTCN: Colloid transport code -- nuclear
This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential-algebraic equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential-algebraic systems
MCAM 5: an advanced interface program for multiple Monte Carlo Codes
The Automatic Modeling Program for Neutronics and Radiation Transport Simulation (MCAM) developed in China, is an advanced interface program between CAD (Computer Aided Design) systems and Monte Carlo (MC) codes. It can significantly reduce the manpower and enhance reliability for constructing MC codes input of complex systems. The latest version MCAM 4.8 was a mature and efficient version which was benchmarked with ITER benchmark model and has been used by hundreds of institutes in more than 40 countries all over the world. It can deal with MCNP and TRIPOLI models. The main function of MCAM is to convert geometries in CAD systems to geometries in MC codes input files. The MCAM version 5.2 is going to be released with added capabilities to support SuperMC, Geant4 and FLUKA Monte Carlo codes
Study on random number generator in Monte Carlo code
The Monte Carlo code uses a sequence of pseudo-random numbers with a random number generator (RNG) to simulate particle histories. A pseudo-random number has its own period depending on its generation method and the period is desired to be long enough not to exceed the period during one Monte Carlo calculation to ensure the correctness especially for a standard deviation of results. The linear congruential generator (LCG) is widely used as Monte Carlo RNG and the period of LCG is not so long by considering the increasing rate of simulation histories in a Monte Carlo calculation according to the remarkable enhancement of computer performance. Recently, many kinds of RNG have been developed and some of their features are better than those of LCG. In this study, we investigate the appropriate RNG in a Monte Carlo code as an alternative to LCG especially for the case of enormous histories. It is found that xorshift has desirable features compared with LCG, and xorshift has a larger period, a comparable speed to generate random numbers, a better randomness, and good applicability to parallel calculation. (author)
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 (2H+) 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 BodyBuilderTM 1.3 anthropomorphic phantoms. Phantoms were modified to allow calculation of the effective dose to a Reference Person using tissues and tissue weighting factors from 1990 and 2007 recommendations of the International Commission on Radiological Protection (ICRP) and gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. Coefficients for the equivalent and effective dose incorporated a radiation weighting factor of 2. At 15 of 19 energies for which coefficients for the effective dose were calculated, coefficients based on ICRP 1990 and 2007 recommendations differed by < 3 %. The greatest difference, 47 %, occurred at 30 MeV. (authors)
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 (3He2+) 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 BodyBuilderTM 1.3 anthropomorphic phantoms modified to allow calculation of effective dose using tissues and tissue weighting factors from either the 1990 or 2007 recommendations of the International Commission on Radiological Protection (ICRP), and gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. At 15 of the 19 energies for which coefficients for effective dose were calculated, coefficients based on ICRP 2007 and 1990 recommendations differed by less than 2%. The greatest difference, 62%, occurred at 100 MeV. Published by Oxford Univ. Press on behalf of the U.S. Government 2010. (authors)
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 (3H+) 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 BodyBuilderTM 1.3 anthropomorphic phantoms. Phantoms were modified to allow calculation of effective dose to a Reference Person using tissues and tissue weighting factors from 1990 and 2007 recommendations of the International Commission on Radiological Protection (ICRP) and calculation of gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. At 15 of the 19 energies for which coefficients for effective dose were calculated, coefficients based on ICRP 2007 and 1990 recommendations differed by less than 3%. The greatest difference, 43%, occurred at 30 MeV. Published by Oxford Univ. Press on behalf of the US Government 2010. (authors)
The techniques of learning theory and pattern recognition are used to learn splitting surface locations for the Monte Carlo neutron transport code MCN. A study is performed to determine default values for several pattern recognition and learning parameters. The modified MCN code is used to reduce computer cost for several nontrivial example problems
Computer codes in nuclear safety, radiation transport and dosimetry
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
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
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.
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.
A code to simulate almost any electron--photon transport problem conceivable is described. The report begins with a lengthy historical introduction and a description of the shower generation process. Then the detailed physics of the shower processes and the methods used to simulate them are presented. Ideas of sampling theory, transport techniques, particle interactions in general, and programing details are discussed. Next, EGS calculations and various experiments and other Monte Carlo results are compared. The remainder of the report consists of user manuals for EGS, PEGS, and TESTSR codes; options, input specifications, and typical output are included. 38 figures, 12 tables
On the use of SERPENT Monte Carlo code to generate few group diffusion constants
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)
A semianalytic Monte Carlo code for modelling LIDAR measurements
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.
Monte Carlo capabilities of the SCALE code system
Highlights: • Foundational Monte Carlo capabilities of SCALE are described. • Improvements in continuous-energy treatments are detailed. • New methods for problem-dependent temperature corrections are described. • New methods for sensitivity analysis and depletion are described. • Nuclear data, users interfaces, and quality assurance activities are summarized. - Abstract: SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a “plug-and-play” framework that includes three deterministic and three Monte Carlo radiation transport solvers that can be selected based on the desired solution, including hybrid deterministic/Monte Carlo simulations. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE’s graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2 will provide several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, and sensitivity and uncertainty analysis. An overview of the Monte Carlo capabilities of SCALE is provided here, with emphasis on new features for SCALE 6.2
The use of an inbuilt importance generator for acceleration of the Monte Carlo code MCBEND
Monte Carlo is currently the most accurate method for the analysis of neutron and gamma-ray transport. However its application, especially to deep penetration studies, is costly in terms of the man-days to set up the calculation and in terms of computer usage. The MAGIC module, developed at the Winfrith Technology Centre, addresses both these problems. It employs an automated procedure based upon the established technique of splitting/roulette with an importance function derived from the solution of the adjoint diffusion equation. Examples are given of the application of the module with Monte Carlo code MCBEND
The macro response Monte Carlo method for electron transport
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
Burnup calculation methodology in the serpent 2 Monte Carlo code
This paper presents two topics related to the burnup calculation capabilities in the Serpent 2 Monte Carlo code: advanced time-integration methods and improved memory management, accomplished by the use of different optimization modes. The development of the introduced methods is an important part of re-writing the Serpent source code, carried out for the purpose of extending the burnup calculation capabilities from 2D assembly-level calculations to large 3D reactor-scale problems. The progress is demonstrated by repeating a PWR test case, originally carried out in 2009 for the validation of the newly-implemented burnup calculation routines in Serpent 1. (authors)
SPAMCART: a code for smoothed particle Monte Carlo radiative transfer
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.
Optimization of Monte Carlo transport simulations in stochastic media
This paper presents an accurate and efficient approach to optimize radiation transport simulations in a stochastic medium of high heterogeneity, like the Very High Temperature Gas-cooled Reactor (VHTR) configurations packed with TRISO fuel particles. Based on a fast nearest neighbor search algorithm, a modified fast Random Sequential Addition (RSA) method is first developed to speed up the generation of the stochastic media systems packed with both mono-sized and poly-sized spheres. A fast neutron tracking method is then developed to optimize the next sphere boundary search in the radiation transport procedure. In order to investigate their accuracy and efficiency, the developed sphere packing and neutron tracking methods are implemented into an in-house continuous energy Monte Carlo code to solve an eigenvalue problem in VHTR unit cells. Comparison with the MCNP benchmark calculations for the same problem indicates that the new methods show considerably higher computational efficiency. (authors)
Simulating fast transients with fuel behavior feedback using the Serpent 2 Monte Carlo code
Simulating transients with reactivity feedback effects using Monte Carlo neutron transport codes can be used for validating deterministic transient codes or estimating for example the total deposited energy in a fuel rod following a known reactivity insertion in the system. Recent increases in computational power as well as developments in calculation methodology makes it possible to obtain a coupled solution for several aspects of the multi-physics problem in a single calculation. This paper describes the different methods implemented in Serpent 2 Monte Carlo code that enable it to model fast transients with fuel behavior feedback. The capability is demonstrated in a prompt critical pin-cell case, where the transient is shut down by the negative reactivity from rising fuel temperature. (author)
ERSN-OpenMC, a Java-based GUI for OpenMC Monte Carlo code
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.
TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticallity studies
TRIMARAN is developed for safety analysis of nuclar components containing fissionnable materials: shipping casks, storage and cooling pools, manufacture and reprocessing plants. It solves the transport equation by Monte Carlo method in general three dimensional geometry with multigroup P1 approximation. A special representation of cross sections and numbers has been developed in order to reduce considerably the computing cost and allow this three dimensional code to compete with standard numerical program used in parametric studies
Development of Monte Carlo decay gamma-ray transport calculation system
In the DT fusion reactor, it is critical concern to evaluate the decay gamma-ray biological dose rates after the reactor shutdown exactly. In order to evaluate the decay gamma-ray biological dose rates exactly, three dimensional Monte Carlo decay gamma-ray transport calculation system have been developed by connecting the three dimensional Monte Carlo particle transport calculation code and the induced activity calculation code. The developed calculation system consists of the following four functions. (1) The operational neutron flux distribution is calculated by the three dimensional Monte Carlo particle transport calculation code. (2) The induced activities are calculated by the induced activity calculation code. (3) The decay gamma-ray source distribution is obtained from the induced activities. (4) The decay gamma-rays are generated by using the decay gamma-ray source distribution, and the decay gamma-ray transport calculation is conducted by the three dimensional Monte Carlo particle transport calculation code. In order to reduce the calculation time drastically, a biasing system for the decay gamma-ray source distribution has been developed, and the function is also included in the present system. In this paper, the outline and the detail of the system, and the execution example are reported. The evaluation for the effect of the biasing system is also reported. (author)
Monte Carlo Capabilities of the SCALE Code System
Rearden, B. T.; Petrie, L. M.; Peplow, D. E.; Bekar, K. B.; Wiarda, D.; Celik, C.; Perfetti, C. M.; Ibrahim, A. M.; Hart, S. W. D.; Dunn, M. E.
2014-06-01
SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a "plug-and-play" framework that includes three deterministic and three Monte Carlo radiation transport solvers that can be selected based on the desired solution, including hybrid deterministic/Monte Carlo simulations. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE's graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2, to be released in 2014, will provide several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, and sensitivity and uncertainty analysis. An overview of the Monte Carlo capabilities of SCALE is provided here, with emphasis on new features for SCALE 6.2.
The present report describes a computer code DEEP which calculates the organ dose equivalents and the effective dose equivalent for external photon exposure by the Monte Carlo method. MORSE-CG, Monte Carlo radiation transport code, is incorporated into the DEEP code to simulate photon transport phenomena in and around a human body. The code treats an anthropomorphic phantom represented by mathematical formulae and user has a choice for the phantom sex: male, female and unisex. The phantom can wear personal dosimeters on it and user can specify their location and dimension. This document includes instruction and sample problem for the code as well as the general description of dose calculation, human phantom and computer code. (author)
Criticality qualification of a new Monte Carlo code for reactor core analysis
In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal-hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the 'Accelerator part' of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the 'Reactor part' of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.
Criticality qualification of a new Monte Carlo code for reactor core analysis
Catsaros, N. [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece); Gaveau, B. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Jaekel, M. [Laboratoire de Physique Theorique, Ecole Normale Superieure, 24 rue Lhomond, 75231 Paris (France); Maillard, J. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); CNRS-IDRIS, Bt 506, BP167, 91403 Orsay (France); CNRS-IN2P3, 3 rue Michel Ange, 75794 Paris (France); Maurel, G. [Faculte de Medecine, Universite Paris VI, 27 rue de Chaligny, 75012 Paris (France); MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Savva, P., E-mail: savvapan@ipta.demokritos.g [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece); Silva, J. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Varvayanni, M.; Zisis, Th. [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece)
2009-11-15
In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal-hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the 'Accelerator part' of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the 'Reactor part' of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.
Neutron spectrum obtained with Monte Carlo and transport theory
The development of the computer, resulting in increasing memory capacity and processing speed, has enabled the application of Monte Carlo method to estimate the fluxes in thousands of fine bin energy structure. Usually the MC calculation is made using continuous energy nuclear data and exact geometry. Self shielding and interference of nuclides resonances are properly considered. Therefore, the fluxes obtained by this method may be a good estimation of the neutron energy distribution (spectrum) for the problem. In an early work it was proposed to use these fluxes as weighting spectrum to generate multigroup cross section for fast reactor analysis using deterministic codes. This non-traditional use of MC calculation needs a validation to gain confidence in the results. The work presented here is the validation start step of this scheme. The spectra of the JOYO first core fuel assembly MK-I and the benchmark Godiva were calculated using the tally flux estimator of the MCNP code and compared with the reference. Also, the two problems were solved with the multigroup transport theory code XSDRN of the AMPX system using the 171 energy groups VITAMIN-C library. The spectra differences arising from the utilization of these codes, the influence of evaluated data file and the application to fast reactor calculation are discussed. (author)
Efficient data management techniques implemented in the Karlsruhe Monte Carlo code KAMCCO
The Karlsruhe Monte Carlo Code KAMCCO is a forward neutron transport code with an eigenfunction and a fixed source option, including time-dependence. A continuous energy model is combined with a detailed representation of neutron cross sections, based on linear interpolation, Breit-Wigner resonances and probability tables. All input is processed into densely packed, dynamically addressed parameter fields and networks of pointers (addresses). Estimation routines are decoupled from random walk and analyze a storage region with sample records. This technique leads to fast execution with moderate storage requirements and without any I/O-operations except in the input and output stages. 7 references. (U.S.)
Nanodosimetric verification in proton therapy: Monte Carlo Codes Comparison
Full text: Nanodosimetry strives to develop a novel dosimetry concept suitable for advanced modalities of cancer radiotherapy, such as proton therapy. This project aims to evaluate the plausibility of the physical models implemented in the Geant4 Very Low Energy (Geant4-DNA) extensions by comparing nanodosimetric quantities calculated with Geant4-DNA and the PTB Monte Carlo track structure code. Nanodosimetric track structure parameters were calculated for cylindrical targets representing DNA and nucleosome segments and converted into the probability of producing a DSB using the model proposed by Garty et al. [1]. Monoenergetic protons and electrons of energies typical for 6-electron spectra were considered as primary particles. Good agreement was found between the two codes for electrons of energies above 200 eV. Below this energy Geant4-DNA produced slightly higher numbers of ionisations in the sensitive volumes and higher probabilities for DSB formation. For protons, Geant4-DNA also gave higher numbers of ionisations and DSB probabilities, particularly in the low energy range, while a satisfactory agreement was found for energies higher than I MeV. Comparing two codes can be useful as any observed divergence in results between the two codes provides valuable information as to where further consideration of the underlying physical models used in each code may be required. Consistently it was seen that the largest difference between the codes was in the low energy ranges for each particle type. (author)
Introduction to the simulation with MCNP Monte Carlo code and its applications in Medical Physics
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)
Status of vectorized Monte Carlo for particle transport analysis
The conventional particle transport Monte Carlo algorithm is ill suited for modern vector supercomputers because the random nature of the particle transport process in the history based algorithm inhibits construction of vectors. An alternative, event-based algorithm is suitable for vectorization and has been used recently to achieve impressive gains in performance on vector supercomputers. This review describes the event-based algorithm and several variations of it. Implementations of this algorithm for applications in particle transport are described, and their relative merits are discussed. The implementation of Monte Carlo methods on multiple vector parallel processors is considered, as is the potential of massively parallel processors for Monte Carlo particle transport simulations
An analytical solution to a simplified EDXRF model for Monte Carlo code verification
The objective of this study is to obtain an analytical solution to the scalar photon transport equation that can be used to obtain benchmark results for the verification of energy dispersive X-Ray fluorescence (EDXRF) Monte Carlo simulation codes. The multi-collided flux method (multiple scattering method) is implemented to obtain analytical expressions for the space-, energy-, and angle-dependent scalar photon flux for a one dimensional EDXRF model problem. In order to obtain benchmark results, higher-order multiple scattering terms are included in the multi-collided flux method. The details of the analytical solution and of the proposed EDXRF model problem are presented. Analytical expressions obtained are then used to calculate the energy-dependent current. The analytically-calculated energy-dependent current is compared with Monte Carlo code results. The findings of this study show that analytical solutions to the scalar photon transport equation with the proposed model problem can be used as a verification tool in EDXRF Monte Carlo code development.
Computed radiography simulation using the Monte Carlo code MCNPX
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. (author)
Adjoint Monte Carlo techniques and codes for organ dose calculations
Adjoint Monte Carlo simulations can be effectively used for the estimation of doses in small targets when the sources are extended in large volumes or surfaces. The main features of two computer codes for calculating doses at free points or in organs of an anthropomorphic phantom are described. In the first program (REBEL-3) natural gamma-emitting sources are contained in the walls of a dwelling room; in the second one (POKER-CAMP) the user can specify arbitrary gamma sources with different spatial distributions in the environment: in (or on the surface of) the ground and in the air. 3 figures
TRIPOLI-4{sup ®} Monte Carlo code ITER A-lite neutronic model validation
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.
Review of the Monte Carlo and deterministic codes in radiation protection and dosimetry
Modelling a physical system can be carried out either stochastically or deterministically. An example of the former method is the Monte Carlo technique, in which statistically approximate methods are applied to exact models. No transport equation is solved as individual particles are simulated and some specific aspect (tally) of their average behaviour is recorded. The average behaviour of the physical system is then inferred using the central limit theorem. In contrast, deterministic codes use mathematically exact methods that are applied to approximate models to solve the transport equation for the average particle behaviour. The physical system is subdivided in boxes in the phase-space system and particles are followed from one box to the next. The smaller the boxes the better the approximations become. Although the Monte Carlo method has been used for centuries, its more recent manifestation has really emerged from the Manhattan project of the Word War II. Its invention is thought to be mainly due to Metropolis, Ulah (through his interest in poker), Fermi, von Neuman and Richtmeyer. Over the last 20 years or so, the Monte Carlo technique has become a powerful tool in radiation transport. This is due to users taking full advantage of richer cross section data, more powerful computers and Monte Carlo techniques for radiation transport, with high quality physics and better known source spectra. This method is a common sense approach to radiation transport and its success and popularity is quite often also due to necessity, because measurements are not always possible or affordable. In the Monte Carlo method, which is inherently realistic because nature is statistical, a more detailed physics is made possible by isolation of events while rather elaborate geometries can be modelled. Provided that the physics is correct, a simulation is exactly analogous to an experimenter counting particles. In contrast to the deterministic approach, however, a disadvantage of the
Development of a space radiation Monte Carlo computer simulation based on the FLUKA and ROOT codes
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...
A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)
Uncertainties associated with the use of the KENO Monte Carlo criticality codes
The KENO multi-group Monte Carlo criticality codes have earned the reputation of being efficient, user friendly tools especially suited for the analysis of situations commonly encountered in the storage and transportation of fissile materials. Throughout their twenty years of service, a continuing effort has been made to maintain and improve these codes to meet the needs of the nuclear criticality safety community. Foremost among these needs is the knowledge of how to utilize the results safely and effectively. Therefore it is important that code users be aware of uncertainties that may affect their results. These uncertainties originate from approximations in the problem data, methods used to process cross sections, and assumptions, limitations and approximations within the criticality computer code itself. 6 refs., 8 figs., 1 tab
Vectorization of continuous energy Monte Carlo code VIM
VIM is a continuous energy Monte Carlo code for criticality calculation. The random walk control system which uses combinatorial geometry system has been vectorized on FACOM VP-100. Vectorization has been done by the event bank method which controls simultaneous multiple particle's random walks, since behavior of neutron is independent. In vectorization of VIM code, we have two problems. One is a large overhead introduced by program modifications for vectorization. Another is a lowering of vector processing efficiency, since the vector length decreases with time according to the absorption and leakage of neutron and cut off of neutron for variance reduction. The average vector length during the random walks has been kept long by utilizing cross section library of single energy band and by reducing the number of the event banks. The performance ratio of vectorized version to the original one is 1.39 for the simple geometry and 1.13 for the complex geometry. (author)
Neutron transport calculations using Quasi-Monte Carlo methods
Moskowitz, B.S.
1997-07-01
This paper examines the use of quasirandom sequences of points in place of pseudorandom points in Monte Carlo neutron transport calculations. For two simple demonstration problems, the root mean square error, computed over a set of repeated runs, is found to be significantly less when quasirandom sequences are used ({open_quotes}Quasi-Monte Carlo Method{close_quotes}) than when a standard Monte Carlo calculation is performed using only pseudorandom points.
The three-dimensional Monte-Carlo code TRIPOLI-02
TRIPOLI-2 solves the transport equation for neutrons or gamma rays in tridimensional geometrical configurations. TRIPOLI uses the Monte Carlo method. This method allows to treat exactly the geometrical configurations, the energy losses and the scattering laws. TRIPOLI 2 allows to treat the following problems: gamma transport problems, neutrons transport problems with fixed source (the problems can be time dependent or not), critical problems without fixed source and research of multiplication factor due to fissions, subcritical problems with fixed source and with multiplication by fission. These problems can be separate in two types. First type: shielding problems essentially with deep penetration and streaming through voids. Biasing technics are used to reduce the computing time. Second type: core problems for cell calculations or for small core calculations. In this case, it is necessary to have a fine representation of the cross sections. The thermalization is also treated exactly
Importance function by collision probabilities for Monte Carlo code Tripoli
We present a completely automatic biasing technique where the parameters of the biased simulation are deduced from the solution of the adjoint transport equation calculated by collision probabilities. In this study we shall estimate the importance function through collision probabilities method and we shall evaluate its possibilities thanks to a Monte Carlo calculation. We have run simulations with this new biasing method for one-group transport problems with isotropic shocks (one dimension geometry and X-Y geometry) and for multigroup problems with anisotropic shocks (one dimension geometry). For the anisotropic problems we solve the adjoint equation with anisotropic collision probabilities. The results show that for the one-group and homogeneous geometry transport problems the method is quite optimal without Splitting and Russian Roulette technique but for the multigroup and heterogeneous X-Y geometry ones the figures of merit are higher if we add Splitting and Russian Roulette technique
Parallel computing by Monte Carlo codes MVP/GMVP
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 parallel computing platforms or by using a standard parallelization library MPI. The platforms used for benchmark calculations 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, IBM SP2. As mentioned generally, linear speedup could be obtained for large-scale problems but parallelization efficiency decreased as the batch size per a processing element(PE) was smaller. It was also found that the statistical uncertainty for assembly powers was less than 0.1% by the PWR full-core calculation with more than 10 million histories and it took about 1.5 hours by massively parallel computing. (author)
The application of Monte Carlo method to electron and photon beams transport
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
Calculation of Gamma-ray Responses for HPGe Detectors with TRIPOLI-4 Monte Carlo Code
Lee, Yi-Kang; Garg, Ruchi
2014-06-01
The gamma-ray response calculation of HPGe (High Purity Germanium) detector is one of the most important topics of the Monte Carlo transport codes for nuclear instrumentation applications. In this study the new options of TRIPOLI-4 Monte Carlo transport code for gamma-ray spectrometry were investigated. Recent improvements include the gamma-rays modeling of the electron-position annihilation, the low energy electron transport modeling, and the low energy characteristic X-ray production. The impact of these improvements on the detector efficiency of the gamma-ray spectrometry calculations was verified. Four models of HPGe detectors and sample sources were studied. The germanium crystal, the dead layer of the crystal, the central hole, the beryllium window, and the metal housing are the essential parts in detector modeling. A point source, a disc source, and a cylindrical extended source containing a liquid radioactive solution were used to study the TRIPOLI-4 calculations for the gamma-ray energy deposition and the gamma-ray self-shielding. The calculations of full-energy-peak and total detector efficiencies for different sample-detector geometries were performed. Using TRIPOLI-4 code, different gamma-ray energies were applied in order to establish the efficiency curves of the HPGe gamma-ray detectors.