Cramer, S.N.
1984-01-01
The MORSE code is a large general-use multigroup Monte Carlo code system. Although no claims can be made regarding its superiority in either theoretical details or Monte Carlo techniques, MORSE has been, since its inception at ORNL in the late 1960s, the most widely used Monte Carlo radiation transport code. The principal reason for this popularity is that MORSE is relatively easy to use, independent of any installation or distribution center, and it can be easily customized to fit almost any specific need. Features of the MORSE code are described.
Monte Carlo Ion Transport Analysis Code.
Energy Science and Technology Software Center (ESTSC)
2009-04-15
Version: 00 TRIPOS is a versatile Monte Carlo ion transport analysis code. It has been applied to the treatment of both surface and bulk radiation effects. The media considered is composed of multilayer polyatomic materials.
Space Applications of the FLUKA Monte-Carlo Code: Lunar and Planetary Exploration
NASA Technical Reports Server (NTRS)
Anderson, V.; Ballarini, F.; Battistoni, G.; Campanella, M.; Carboni, M.; Cerutti, F.; Elkhayari, N.; Empl, A.; Fasso, A.; Ferrari, A.; Gadoli, E.; Gazelli, M. V.; LeBourgeois, M.; Lee, K. T.; Mayes, B.; Muraro, S.; Ottolenghi, A.; Pelliccioni, M.; Pinsky, L. S.; Rancati, T.; Ranft, J.; Roesler, S.; Sala, P. R.; Scannocchio, D.; Smirnov, G.
2004-01-01
NASA has recognized the need for making additional heavy-ion collision measurements at the U.S. Brookhaven National Laboratory in order to support further improvement of several particle physics transport-code models for space exploration applications. FLUKA has been identified as one of these codes and we will review the nature and status of this investigation as it relates to high-energy heavy-ion physics.
MCMAC: Monte Carlo Merger Analysis Code
NASA Astrophysics Data System (ADS)
Dawson, William A.
2014-07-01
Monte Carlo Merger Analysis Code (MCMAC) aids in the study of merging clusters. It takes observed priors on each subcluster's mass, radial velocity, and projected separation, draws randomly from those priors, and uses them in a analytic model to get posterior PDF's for merger dynamic properties of interest (e.g. collision velocity, time since collision).
THE MCNPX MONTE CARLO RADIATION TRANSPORT CODE
WATERS, LAURIE S.; MCKINNEY, GREGG W.; DURKEE, JOE W.; FENSIN, MICHAEL L.; JAMES, MICHAEL R.; JOHNS, RUSSELL C.; PELOWITZ, DENISE B.
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.
1983-02-01
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. (WHK)
Monte Carlo simulation of turnover processes in the lunar regolith
NASA Technical Reports Server (NTRS)
Arnold, J. R.
1975-01-01
A Monte Carlo model for the gardening of the lunar surface by meteoritic impact is described, and some representative results are given. The model accounts with reasonable success for a wide variety of properties of the regolith. The smoothness of the lunar surface on a scale of centimeters to meters, which was not reproduced in an earlier version of the model, is accounted for by the preferential downward movement of low-energy secondary particles. The time scale for filling lunar grooves and craters by this process is also derived. The experimental bombardment ages (about 4 x 10 to the 8th yr for spallogenic rare gases, about 10 to the 9th yr for neutron capture Gd and Sm isotopes) are not reproduced by the model. The explanation is not obvious.
SPQR: a Monte Carlo reactor kinetics code. [LMFBR
Cramer, S.N.; Dodds, H.L.
1980-02-01
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.
Catastrophic rupture of lunar rocks - A Monte Carlo simulation
NASA Technical Reports Server (NTRS)
Hoerz, F.; Schneider, E.; Gault, D. E.; Hartung, J. B.; Brownlee, D. E.
1975-01-01
A computer model based on Monte Carlo techniques was developed to simulate the destruction of lunar rocks by 'catastrophic rupture' due to meteoroid impact. Energies necessary to accomplish catastrophic rupture were derived from laboratory experiments. A crater-production rate derived from lunar rocks was utilized to calculate absolute time scales. Calculated median survival times for crystalline lunar rocks are 1.9, 4.6, 10.3, and 22 m.y. for rock masses of 10, 100, 1000, and 10,000 g, respectively. Corresponding times of 6, 14.5, 32, and 68 million years are required before the probability of destruction reaches 0.99. These results are consistent with absolute exposure ages measured on returned rocks. Some results also substantiate previous conclusions that the catastrophic-rupture process is significantly more effective in obliterating lunar rocks than mass wasting by single-particle abrasion. The view is also corroborated that most rocks presently on the lunar surface either are exhumed from the regolith or are fragments of much larger boulders rather than primary ejecta excavated from pristine bedrock.
Monte Carlo Model Insights into the Lunar Sodium Exosphere
NASA Technical Reports Server (NTRS)
Hurley, Dana M.; Killen, R. M.; Sarantos, M.
2012-01-01
Sodium in the lunar exosphere is released from the lunar regolith by several mechanisms. These mechanisms include photon stimulated desorption (PSD), impact vaporization, electron stimulated desorption, and ion sputtering. Usually, PSD dominates; however, transient events can temporarily enhance other release mechanisms so that they are dominant. Examples of transient events include meteor showers and coronal mass ejections. The interaction between sodium and the regolith is important in determining the density and spatial distribution of sodium in the lunar exosphere. The temperature at which sodium sticks to the surface is one factor. In addition, the amount of thermal accommodation during the encounter between the sodium atom and the surface affects the exospheric distribution. Finally, the fraction of particles that are stuck when the surface is cold that are rereleased when the surface warms up also affects the exospheric density. In [1], we showed the "ambient" sodium exosphere from Monte Carlo modeling with a fixed source rate and fixed surface interaction parameters. We compared the enhancement when a CME passes the Moon to the ambient conditions. Here, we compare model results to data in order to determine the source rates and surface interaction parameters that provide the best fit of the model to the data.
ALEPH2 - A general purpose Monte Carlo depletion code
Stankovskiy, A.; Van Den Eynde, G.; Baeten, P.; Trakas, C.; Demy, P. M.; Villatte, L.
2012-07-01
The Monte-Carlo burn-up code ALEPH is being developed at SCK-CEN since 2004. A previous version of the code implemented the coupling between the Monte Carlo transport (any version of MCNP or MCNPX) and the ' deterministic' depletion code ORIGEN-2.2 but had important deficiencies in nuclear data treatment and limitations inherent to ORIGEN-2.2. A new version of the code, ALEPH2, has several unique features making it outstanding among other depletion codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. The last generation general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII and JENDL-4) are fully implemented, including special purpose activation, spontaneous fission, fission product yield and radioactive decay data. The built-in depletion algorithm allows to eliminate the uncertainties associated with obtaining the time-dependent nuclide concentrations. A predictor-corrector mechanism, calculation of nuclear heating, calculation of decay heat, decay neutron sources are available as well. The validation of the code on the results of REBUS experimental program has been performed. The ALEPH2 has shown better agreement with measured data than other depletion codes. (authors)
Recent advances in the Mercury Monte Carlo particle transport code
Brantley, P. S.; Dawson, S. A.; McKinley, M. S.; O'Brien, M. J.; Stevens, D. E.; Beck, B. R.; Jurgenson, E. D.; Ebbers, C. A.; Hall, J. M.
2013-07-01
We review recent physics and computational science advances in the Mercury Monte Carlo particle transport code under development at Lawrence Livermore National Laboratory. We describe recent efforts to enable a nuclear resonance fluorescence capability in the Mercury photon transport. We also describe recent work to implement a probability of extinction capability into Mercury. We review the results of current parallel scaling and threading efforts that enable the code to run on millions of MPI processes. (authors)
Monte Carlo Nucleon Meson Transport Code System.
Energy Science and Technology Software Center (ESTSC)
2000-11-17
Version 00 NMTC/JAERI97 is an upgraded version of the code system NMTC/JAERI, which was developed in 1982 at JAERI and is based on the CCC-161/NMTC code system. NMTC/JAERI97 simulates high energy nuclear reactions and nucleon-meson transport processes.
A semianalytic Monte Carlo code for modelling LIDAR measurements
NASA Astrophysics Data System (ADS)
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.
Improved diffusion coefficients generated from Monte Carlo codes
Herman, B. R.; Forget, B.; Smith, K.; Aviles, B. N.
2013-07-01
Monte Carlo codes are becoming more widely used for reactor analysis. Some of these applications involve the generation of diffusion theory parameters including macroscopic cross sections and diffusion coefficients. Two approximations used to generate diffusion coefficients are assessed using the Monte Carlo code MC21. The first is the method of homogenization; whether to weight either fine-group transport cross sections or fine-group diffusion coefficients when collapsing to few-group diffusion coefficients. The second is a fundamental approximation made to the energy-dependent P1 equations to derive the energy-dependent diffusion equations. Standard Monte Carlo codes usually generate a flux-weighted transport cross section with no correction to the diffusion approximation. Results indicate that this causes noticeable tilting in reconstructed pin powers in simple test lattices with L2 norm error of 3.6%. This error is reduced significantly to 0.27% when weighting fine-group diffusion coefficients by the flux and applying a correction to the diffusion approximation. Noticeable tilting in reconstructed fluxes and pin powers was reduced when applying these corrections. (authors)
Current status of the PSG Monte Carlo neutron transport code
Leppaenen, J.
2006-07-01
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)
Burnup calculation methodology in the serpent 2 Monte Carlo code
Leppaenen, J.; Isotalo, A.
2012-07-01
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)
A Post-Monte-Carlo Sensitivity Analysis Code
Energy Science and Technology Software Center (ESTSC)
2000-04-04
SATOOL (Sensitivity Analysis TOOL) is a code for sensitivity analysis, following an uncertainity analysis with Monte Carlo simulations. Sensitivity analysis identifies those input variables, whose variance contributes dominatly to the variance in the output. This analysis can be used to reduce the variance in the output variables by redefining the "sensitive" variables with greater precision, i.e. with lower variance. The code identifies a group of sensitive variables, ranks them in the order of importance andmore » also quantifies the relative importance among the sensitive variables.« less
Computational radiology and imaging with the MCNP Monte Carlo code
Estes, G.P.; Taylor, W.M.
1995-05-01
MCNP, a 3D coupled neutron/photon/electron Monte Carlo radiation transport code, is currently used in medical applications such as cancer radiation treatment planning, interpretation of diagnostic radiation images, and treatment beam optimization. This paper will discuss MCNP`s current uses and capabilities, as well as envisioned improvements that would further enhance MCNP role in computational medicine. It will be demonstrated that the methodology exists to simulate medical images (e.g. SPECT). Techniques will be discussed that would enable the construction of 3D computational geometry models of individual patients for use in patient-specific studies that would improve the quality of care for patients.
Acceleration of a Monte Carlo radiation transport code
Hochstedler, R.D.; Smith, L.M.
1996-03-01
Execution time for the Integrated TIGER Series (ITS) Monte Carlo radiation transport code has been reduced by careful re-coding of computationally intensive subroutines. Three test cases for the TIGER (1-D slab geometry), CYLTRAN (2-D cylindrical geometry), and ACCEPT (3-D arbitrary geometry) codes were identified and used to benchmark and profile program execution. Based upon these results, sixteen top time-consuming subroutines were examined and nine of them modified to accelerate computations with equivalent numerical output to the original. The results obtained via this study indicate that speedup factors of 1.90 for the TIGER code, 1.67 for the CYLTRAN code, and 1.11 for the ACCEPT code are achievable. {copyright} {ital 1996 American Institute of Physics.}
Applications guide to the MORSE Monte Carlo code
Cramer, S.N.
1985-08-01
A practical guide for the implementation of the MORESE-CG Monte Carlo radiation transport computer code system is presented. The various versions of the MORSE code are compared and contrasted, and the many references dealing explicitly with the MORSE-CG code are reviewed. The treatment of angular scattering is discussed, and procedures for obtaining increased differentiality of results in terms of reaction types and nuclides from a multigroup Monte Carlo code are explained in terms of cross-section and geometry data manipulation. Examples of standard cross-section data input and output are shown. Many other features of the code system are also reviewed, including (1) the concept of primary and secondary particles, (2) fission neutron generation, (3) albedo data capability, (4) DOMINO coupling, (5) history file use for post-processing of results, (6) adjoint mode operation, (7) variance reduction, and (8) input/output. In addition, examples of the combinatorial geometry are given, and the new array of arrays geometry feature (MARS) and its three-dimensional plotting code (JUNEBUG) are presented. Realistic examples of user routines for source, estimation, path-length stretching, and cross-section data manipulation are given. A deatiled explanation of the coupling between the random walk and estimation procedure is given in terms of both code parameters and physical analogies. The operation of the code in the adjoint mode is covered extensively. The basic concepts of adjoint theory and dimensionality are discussed and examples of adjoint source and estimator user routines are given for all common situations. Adjoint source normalization is explained, a few sample problems are given, and the concept of obtaining forward differential results from adjoint calculations is covered. Finally, the documentation of the standard MORSE-CG sample problem package is reviewed and on-going and future work is discussed.
Monte-Carlo Continuous Energy Burnup Code System.
Energy Science and Technology Software Center (ESTSC)
2007-08-31
Version 00 MCB is a Monte Carlo Continuous Energy Burnup Code for a general-purpose use to calculate a nuclide density time evolution with burnup or decay. It includes eigenvalue calculations of critical and subcritical systems as well as neutron transport calculations in fixed source mode or k-code mode to obtain reaction rates and energy deposition that are necessary for burnup calculations. The MCB-1C patch file and data packages as distributed by the NEADB are verymore » well organized and are being made available through RSICC as received. The RSICC package includes the MCB-1C patch and MCB data libraries. Installation of MCB requires MCNP4C source code and utility programs, which are not included in this MCB distribution. They were provided with the now obsolete CCC-700/MCNP-4C package.« less
Radiographic Capabilities of the MERCURY Monte Carlo Code
McKinley, M S; von Wittenau, A
2008-04-07
MERCURY is a modern, parallel, general-purpose Monte Carlo code being developed at the Lawrence Livermore National Laboratory. Recently, a radiographic capability has been added. MERCURY can create a source of diagnostic, virtual particles that are aimed at pixels in an image tally. This new feature is compared to the radiography code, HADES, for verification and timing. Comparisons for accuracy were made using the French Test Object and for timing were made by tracking through an unstructured mesh. In addition, self consistency tests were run in MERCURY for the British Test Object and scattering test problem. MERCURY and HADES were found to agree to the precision of the input data. HADES appears to run around eight times faster than the MERCURY in the timing study. Profiling the MERCURY code has turned up several differences in the algorithms which account for this. These differences will be addressed in a future release of MERCURY.
Parallelization of KENO-Va Monte Carlo code
NASA Astrophysics Data System (ADS)
Ramón, Javier; Peña, Jorge
1995-07-01
KENO-Va is a code integrated within the SCALE system developed by Oak Ridge that solves the transport equation through the Monte Carlo Method. It is being used at the Consejo de Seguridad Nuclear (CSN) to perform criticality calculations for fuel storage pools and shipping casks. Two parallel versions of the code: one for shared memory machines and other for distributed memory systems using the message-passing interface PVM have been generated. In both versions the neutrons of each generation are tracked in parallel. In order to preserve the reproducibility of the results in both versions, advanced seeds for random numbers were used. The CONVEX C3440 with four processors and shared memory at CSN was used to implement the shared memory version. A FDDI network of 6 HP9000/735 was employed to implement the message-passing version using proprietary PVM. The speedup obtained was 3.6 in both cases.
New features of the Monte-Carlo code MOCADI
NASA Astrophysics Data System (ADS)
Iwasa, N.; Weick, H.; Geissel, H.
2011-04-01
MOCADI, the Monte-Carlo code for tracking of ions in ion-optical systems with non-Liouvillian elements, has been extended. Accurate atomic and nuclear interactions are taken into account when ions penetrate gaseous and solid matter placed within the ion-optical system. The new features of MOCADI are described in this article with practical examples which demonstrate the new possibilities, such as new event-generators for targets and spontaneous nuclear decay, the option of atomic-charge state fluctuation in matter, loops for multi-turn ion-optical systems and a graphical user interface for easier operating and control of the program. Experiments for investigation of nuclear structure and reactions with ions circulating in a storage ring can now be ideally studied with MOCADI.
RMC - A Monte Carlo code for reactor physics analysis
Wang, K.; Li, Z.; She, D.; Liang, J.; Xu, Q.; Qiu, A.; Yu, J.; Sun, J.; Fan, X.; Yu, G.
2013-07-01
A new Monte Carlo neutron transport code RMC has been being developed by Department of Engineering Physics, Tsinghua University, Beijing as a tool for reactor physics analysis on high-performance computing platforms. To meet the requirements of reactor analysis, RMC now has such functions as criticality calculation, fixed-source calculation, burnup calculation and kinetics simulations. Some techniques for geometry treatment, new burnup algorithm, source convergence acceleration, massive tally and parallel calculation, and temperature dependent cross sections processing are researched and implemented in RMC to improve the efficiency. Validation results of criticality calculation, burnup calculation, source convergence acceleration, tallies performance and parallel performance shown in this paper prove the capabilities of RMC in dealing with reactor analysis problems with good performances. (authors)
FZ2MC: A Tool for Monte Carlo Transport Code Geometry Manipulation
Hackel, B M; Nielsen Jr., D E; Procassini, R J
2009-02-25
The process of creating and validating combinatorial geometry representations of complex systems for use in Monte Carlo transport simulations can be both time consuming and error prone. To simplify this process, a tool has been developed which employs extensions of the Form-Z commercial solid modeling tool. The resultant FZ2MC (Form-Z to Monte Carlo) tool permits users to create, modify and validate Monte Carlo geometry and material composition input data. Plugin modules that export this data to an input file, as well as parse data from existing input files, have been developed for several Monte Carlo codes. The FZ2MC tool is envisioned as a 'universal' tool for the manipulation of Monte Carlo geometry and material data. To this end, collaboration on the development of plug-in modules for additional Monte Carlo codes is desired.
abcpmc: Approximate Bayesian Computation for Population Monte-Carlo code
NASA Astrophysics Data System (ADS)
Akeret, Joel
2015-04-01
abcpmc is a Python Approximate Bayesian Computing (ABC) Population Monte Carlo (PMC) implementation based on Sequential Monte Carlo (SMC) with Particle Filtering techniques. It is extendable with k-nearest neighbour (KNN) or optimal local covariance matrix (OLCM) pertubation kernels and has built-in support for massively parallelized sampling on a cluster using MPI.
Parallelization of a Monte Carlo particle transport simulation code
NASA Astrophysics Data System (ADS)
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.
An Automated, Multi-Step Monte Carlo Burnup Code System.
TRELLUE, HOLLY R.
2003-07-14
Version 02 MONTEBURNS Version 2 calculates coupled neutronic/isotopic results for nuclear systems and produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. MONTEBURNS is a fully automated tool that links the LANL MCNP Monte Carlo transport code with a radioactive decay and burnup code. Highlights on changes to Version 2 are listed in the transmittal letter. Along with other minor improvements in MONTEBURNS Version 2, the option was added to use CINDER90 instead of ORIGEN2 as the depletion/decay part of the system. CINDER90 is a multi-group depletion code developed at LANL and is not currently available from RSICC. This MONTEBURNS release was tested with various combinations of CCC-715/MCNPX 2.4.0, CCC-710/MCNP5, CCC-700/MCNP4C, CCC-371/ORIGEN2.2, ORIGEN2.1 and CINDER90. Perl is required software and is not included in this distribution. MCNP, ORIGEN2, and CINDER90 are not included.
An Automated, Multi-Step Monte Carlo Burnup Code System.
Energy Science and Technology Software Center (ESTSC)
2003-07-14
Version 02 MONTEBURNS Version 2 calculates coupled neutronic/isotopic results for nuclear systems and produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. MONTEBURNS is a fully automated tool that links the LANL MCNP Monte Carlo transport code with a radioactive decay and burnup code. Highlights on changes to Version 2 are listed in the transmittal letter. Along with other minor improvements in MONTEBURNS Version 2,more » the option was added to use CINDER90 instead of ORIGEN2 as the depletion/decay part of the system. CINDER90 is a multi-group depletion code developed at LANL and is not currently available from RSICC. This MONTEBURNS release was tested with various combinations of CCC-715/MCNPX 2.4.0, CCC-710/MCNP5, CCC-700/MCNP4C, CCC-371/ORIGEN2.2, ORIGEN2.1 and CINDER90. Perl is required software and is not included in this distribution. MCNP, ORIGEN2, and CINDER90 are not included.« less
Monte Carlo code for high spatial resolution ocean color simulations.
D'Alimonte, Davide; Zibordi, Giuseppe; Kajiyama, Tamito; Cunha, José C
2010-09-10
A Monte Carlo code for ocean color simulations has been developed to model in-water radiometric fields of downward and upward irradiance (E(d) and E(u)), and upwelling radiance (L(u)) in a two-dimensional domain with a high spatial resolution. The efficiency of the code has been optimized by applying state-of-the-art computing solutions, while the accuracy of simulation results has been quantified through benchmark with the widely used Hydrolight code for various values of seawater inherent optical properties and different illumination conditions. Considering a seawater single scattering albedo of 0.9, as well as surface waves of 5 m width and 0.5 m height, the study has shown that the number of photons required to quantify uncertainties induced by wave focusing effects on E(d), E(u), and L(u) data products is of the order of 10(6), 10(9), and 10(10), respectively. On this basis, the effects of sea-surface geometries on radiometric quantities have been investigated for different surface gravity waves. Data products from simulated radiometric profiles have finally been analyzed as a function of the deployment speed and sampling frequency of current free-fall systems in view of providing recommendations to improve measurement protocols. PMID:20830183
Status of the MORSE multigroup Monte Carlo radiation transport code
Emmett, M.B.
1993-06-01
There are two versions of the MORSE multigroup Monte Carlo radiation transport computer code system at Oak Ridge National Laboratory. MORSE-CGA is the most well-known and has undergone extensive use for many years. MORSE-SGC was originally developed in about 1980 in order to restructure the cross-section handling and thereby save storage. However, with the advent of new computer systems having much larger storage capacity, that aspect of SGC has become unnecessary. Both versions use data from multigroup cross-section libraries, although in somewhat different formats. MORSE-SGC is the version of MORSE that is part of the SCALE system, but it can also be run stand-alone. Both CGA and SGC use the Multiple Array System (MARS) geometry package. In the last six months the main focus of the work on these two versions has been on making them operational on workstations, in particular, the IBM RISC 6000 family. A new version of SCALE for workstations is being released to the Radiation Shielding Information Center (RSIC). MORSE-CGA, Version 2.0, is also being released to RSIC. Both SGC and CGA have undergone other revisions recently. This paper reports on the current status of the MORSE code system.
NASA Technical Reports Server (NTRS)
Firstenberg, H.
1971-01-01
The statistics are considered of the Monte Carlo method relative to the interpretation of the NUGAM2 and NUGAM3 computer code results. A numerical experiment using the NUGAM2 code is presented and the results are statistically interpreted.
The Monte Carlo code MCSHAPE: Main features and recent developments
NASA Astrophysics Data System (ADS)
Scot, Viviana; Fernandez, Jorge E.
2015-06-01
MCSHAPE is a general purpose Monte Carlo code developed at the University of Bologna to simulate the diffusion of X- and gamma-ray photons with the special feature of describing the full evolution of the photon polarization state along the interactions with the target. The prevailing photon-matter interactions in the energy range 1-1000 keV, Compton and Rayleigh scattering and photoelectric effect, are considered. All the parameters that characterize the photon transport can be suitably defined: (i) the source intensity, (ii) its full polarization state as a function of energy, (iii) the number of collisions, and (iv) the energy interval and resolution of the simulation. It is possible to visualize the results for selected groups of interactions. MCSHAPE simulates the propagation in heterogeneous media of polarized photons (from synchrotron sources) or of partially polarized sources (from X-ray tubes). In this paper, the main features of MCSHAPE are illustrated with some examples and a comparison with experimental data.
A 3DHZETRN Code in a Spherical Uniform Sphere with Monte Carlo Verification
NASA Technical Reports Server (NTRS)
Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.
2014-01-01
The computationally efficient HZETRN code has been used in recent trade studies for lunar and Martian exploration and is currently being used in the engineering development of the next generation of space vehicles, habitats, and extra vehicular activity equipment. A new version (3DHZETRN) capable of transporting High charge (Z) and Energy (HZE) and light ions (including neutrons) under space-like boundary conditions with enhanced neutron and light ion propagation is under development. In the present report, new algorithms for light ion and neutron propagation with well-defined convergence criteria in 3D objects is developed and tested against Monte Carlo simulations to verify the solution methodology. The code will be available through the software system, OLTARIS, for shield design and validation and provides a basis for personal computer software capable of space shield analysis and optimization.
Generation of SFR few-group constants using the Monte Carlo code Serpent
Fridman, E.; Rachamin, R.; Shwageraus, E.
2013-07-01
In this study, the Serpent Monte Carlo code was used as a tool for preparation of homogenized few-group cross sections for the nodal diffusion analysis of Sodium cooled Fast Reactor (SFR) cores. Few-group constants for two reference SFR cores were generated by Serpent and then employed by nodal diffusion code DYN3D in 2D full core calculations. The DYN3D results were verified against the references full core Serpent Monte Carlo solutions. A good agreement between the reference Monte Carlo and nodal diffusion results was observed demonstrating the feasibility of using Serpent for generation of few-group constants for the deterministic SFR analysis. (authors)
Monte Carlo Capabilities of the SCALE Code System
NASA Astrophysics Data System (ADS)
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.
Monte Carlo capabilities of the SCALE code system
Rearden, Bradley T.; Petrie, Jr., Lester M.; Peplow, Douglas E.; Bekar, Kursat B.; Wiarda, Dorothea; Celik, Cihangir; Perfetti, Christopher M.; Ibrahim, Ahmad M.; Hart, S. W. D.; Dunn, Michael E.; et al
2014-09-12
SCALE is a broadly 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 asmore » 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. Finally, an overview of the Monte Carlo capabilities of SCALE is provided here, with emphasis on new features for SCALE 6.2.« less
Monte Carlo capabilities of the SCALE code system
Rearden, Bradley T.; Petrie, Jr., Lester M.; Peplow, Douglas E.; Bekar, Kursat B.; Wiarda, Dorothea; Celik, Cihangir; Perfetti, Christopher M.; Ibrahim, Ahmad M.; Hart, S. W. D.; Dunn, Michael E.; Marshall, William J.
2014-09-12
SCALE is a broadly 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. Finally, an overview of the Monte Carlo capabilities of SCALE is provided here, with emphasis on new features for SCALE 6.2.
Multiparticle Monte Carlo Code System for Shielding and Criticality Use.
2015-06-01
Version 00 COG is a modern, full-featured Monte Carlo radiation transport code that provides accurate answers to complex shielding, criticality, and activation problems.COG was written to be state-of-the-art and free of physics approximations and compromises found in earlier codes. COG is fully 3-D, uses point-wise cross sections and exact angular scattering, and allows a full range of biasing options to speed up solutions for deep penetration problems. Additionally, a criticality option is available for computing Keff for assemblies of fissile materials. ENDL or ENDFB cross section libraries may be used. COG home page: http://cog.llnl.gov. Cross section libraries are included in the package. COG can use either the LLNL ENDL-90 cross section set or the ENDFB/VI set. Analytic surfaces are used to describe geometric boundaries. Parts (volumes) are described by a method of Constructive Solid Geometry. Surface types include surfaces of up to fourth order, and pseudo-surfaces such as boxes, finite cylinders, and figures of revolution. Repeated assemblies need be defined only once. Parts are visualized in cross-section and perspective picture views. A lattice feature simplifies the specification of regular arrays of parts. Parallel processing under MPI is supported for multi-CPU systems. Source and random-walk biasing techniques may be selected to improve solution statistics. These include source angular biasing, importance weighting, particle splitting and Russian roulette, pathlength stretching, point detectors, scattered direction biasing, and forced collisions. Criticality For a fissioning system, COG will compute Keff by transporting batches of neutrons through the system. Activation COG can compute gamma-ray doses due to neutron-activated materials, starting with just a neutron source. Coupled Problems COG can solve coupled problems involving neutrons, photons, and electrons. COG 11.1 is an updated version of COG11.1 BETA 2 (RSICC C00777MNYCP02). New
Multiparticle Monte Carlo Code System for Shielding and Criticality Use.
Energy Science and Technology Software Center (ESTSC)
2015-06-01
Version 00 COG is a modern, full-featured Monte Carlo radiation transport code that provides accurate answers to complex shielding, criticality, and activation problems.COG was written to be state-of-the-art and free of physics approximations and compromises found in earlier codes. COG is fully 3-D, uses point-wise cross sections and exact angular scattering, and allows a full range of biasing options to speed up solutions for deep penetration problems. Additionally, a criticality option is available for computingmore » Keff for assemblies of fissile materials. ENDL or ENDFB cross section libraries may be used. COG home page: http://cog.llnl.gov. Cross section libraries are included in the package. COG can use either the LLNL ENDL-90 cross section set or the ENDFB/VI set. Analytic surfaces are used to describe geometric boundaries. Parts (volumes) are described by a method of Constructive Solid Geometry. Surface types include surfaces of up to fourth order, and pseudo-surfaces such as boxes, finite cylinders, and figures of revolution. Repeated assemblies need be defined only once. Parts are visualized in cross-section and perspective picture views. A lattice feature simplifies the specification of regular arrays of parts. Parallel processing under MPI is supported for multi-CPU systems. Source and random-walk biasing techniques may be selected to improve solution statistics. These include source angular biasing, importance weighting, particle splitting and Russian roulette, pathlength stretching, point detectors, scattered direction biasing, and forced collisions. Criticality For a fissioning system, COG will compute Keff by transporting batches of neutrons through the system. Activation COG can compute gamma-ray doses due to neutron-activated materials, starting with just a neutron source. Coupled Problems COG can solve coupled problems involving neutrons, photons, and electrons. COG 11.1 is an updated version of COG11.1 BETA 2 (RSICC C00777MNYCP02
Development of a space radiation Monte Carlo computer simulation based on the FLUKA and ROOT codes.
Pinsky, L S; Wilson, T 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 useful in the design and analysis of experiments such as ACCESS (Advanced Cosmic-ray Composition Experiment for Space Station), which is an Office of Space Science payload currently under evaluation for deployment on the International Space Station (ISS). FLUKA will be significantly improved and tailored for use in simulating space radiation in four ways. First, the additional physics not presently within the code that is necessary to simulate the problems of interest, namely the heavy ion inelastic processes, will be incorporated. Second, the internal geometry package will be replaced with one that will substantially increase the calculation speed as well as simplify the data input task. Third, default incident flux packages that include all of the different space radiation sources of interest will be included. Finally, the user interface and internal data structure will be melded together with ROOT, the object-oriented data analysis infrastructure system. Beyond
Full 3D visualization tool-kit for Monte Carlo and deterministic transport codes
Frambati, S.; Frignani, M.
2012-07-01
We propose a package of tools capable of translating the geometric inputs and outputs of many Monte Carlo and deterministic radiation transport codes into open source file formats. These tools are aimed at bridging the gap between trusted, widely-used radiation analysis codes and very powerful, more recent and commonly used visualization software, thus supporting the design process and helping with shielding optimization. Three main lines of development were followed: mesh-based analysis of Monte Carlo codes, mesh-based analysis of deterministic codes and Monte Carlo surface meshing. The developed kit is considered a powerful and cost-effective tool in the computer-aided design for radiation transport code users of the nuclear world, and in particular in the fields of core design and radiation analysis. (authors)
Review of Fast Monte Carlo Codes for Dose Calculation in Radiation Therapy Treatment Planning
Jabbari, Keyvan
2011-01-01
An important requirement in radiation therapy is a fast and accurate treatment planning system. This system, using computed tomography (CT) data, direction, and characteristics of the beam, calculates the dose at all points of the patient's volume. The two main factors in treatment planning system are accuracy and speed. According to these factors, various generations of treatment planning systems are developed. This article is a review of the Fast Monte Carlo treatment planning algorithms, which are accurate and fast at the same time. The Monte Carlo techniques are based on the transport of each individual particle (e.g., photon or electron) in the tissue. The transport of the particle is done using the physics of the interaction of the particles with matter. Other techniques transport the particles as a group. For a typical dose calculation in radiation therapy the code has to transport several millions particles, which take a few hours, therefore, the Monte Carlo techniques are accurate, but slow for clinical use. In recent years, with the development of the ‘fast’ Monte Carlo systems, one is able to perform dose calculation in a reasonable time for clinical use. The acceptable time for dose calculation is in the range of one minute. There is currently a growing interest in the fast Monte Carlo treatment planning systems and there are many commercial treatment planning systems that perform dose calculation in radiation therapy based on the Monte Carlo technique. PMID:22606661
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.
3D Direct Simulation Monte Carlo Code Which Solves for Geometrics
Energy Science and Technology Software Center (ESTSC)
1998-01-13
Pegasus is a 3D Direct Simulation Monte Carlo Code which solves for geometries which can be represented by bodies of revolution. Included are all the surface chemistry enhancements in the 2D code Icarus as well as a real vacuum pump model. The code includes multiple species transport.
PEGASUS. 3D Direct Simulation Monte Carlo Code Which Solves for Geometrics
Bartel, T.J.
1998-12-01
Pegasus is a 3D Direct Simulation Monte Carlo Code which solves for geometries which can be represented by bodies of revolution. Included are all the surface chemistry enhancements in the 2D code Icarus as well as a real vacuum pump model. The code includes multiple species transport.
FREYA-a new Monte Carlo code for improved modeling of fission chains
Hagmann, C A; Randrup, J; Vogt, R L
2012-06-12
A new simulation capability for modeling of individual fission events and chains and the transport of fission products in materials is presented. FREYA ( Fission Yield Event Yield Algorithm ) is a Monte Carlo code for generating fission events providing correlated kinematic information for prompt neutrons, gammas, and fragments. As a standalone code, FREYA calculates quantities such as multiplicity-energy, angular, and gamma-neutron energy sharing correlations. To study materials with multiplication, shielding effects, and detectors, we have integrated FREYA into the general purpose Monte Carlo code MCNP. This new tool will allow more accurate modeling of detector responses including correlations and the development of SNM detectors with increased sensitivity.
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
HTGR Reactor Physics and Burnup Calculations Using the Serpent Monte Carlo Code
Leppanen, Jaakko; DeHart, Mark D
2009-01-01
One of the main advantages of the continuous-energy Monte Carlo method is its versatility and the capability to model any fuel or reactor configuration without major approximations. This capability becomes particularly valuable in studies involving innovative reactor designs and next-generation systems, which often lie beyond the capabilities of deterministic LWR transport codes. In this study, a conceptual prismatic HTGR fuel assembly was modeled using the Serpent Monte Carlo reactor physics burnup calculation code, under development at VTT Technical Research Centre of Finland since 2004. A new explicit particle fuel model was developed to account for the heterogeneity effects. The results are compared to other Monte Carlo and deterministic transport codes and the study also serves as a test case for the modules and methods in SCALE 6.
T.J. Urbatsch; T.M. Evans
2006-02-15
We have released Version 2 of Milagro, an object-oriented, C++ code that performs radiative transfer using Fleck and Cummings' Implicit Monte Carlo method. Milagro, a part of the Jayenne program, is a stand-alone driver code used as a methods research vehicle and to verify its underlying classes. These underlying classes are used to construct Implicit Monte Carlo packages for external customers. Milagro-2 represents a design overhaul that allows better parallelism and extensibility. New features in Milagro-2 include verified momentum deposition, restart capability, graphics capability, exact energy conservation, and improved load balancing and parallel efficiency. A users' guide also describes how to configure, make, and run Milagro2.
Revised methods for few-group cross sections generation in the Serpent Monte Carlo code
Fridman, E.; Leppaenen, J.
2012-07-01
This paper presents new calculation methods, recently implemented in the Serpent Monte Carlo code, and related to the production of homogenized few-group constants for deterministic 3D core analysis. The new methods fall under three topics: 1) Improved treatment of neutron-multiplying scattering reactions, 2) Group constant generation in reflectors and other non-fissile regions and 3) Homogenization in leakage-corrected criticality spectrum. The methodology is demonstrated by a numerical example, comparing a deterministic nodal diffusion calculation using Serpent-generated cross sections to a reference full-core Monte Carlo simulation. It is concluded that the new methodology improves the results of the deterministic calculation, and paves the way for Monte Carlo based group constant generation. (authors)
Improved methods of handling massive tallies in reactor Monte Carlo Code RMC
She, D.; Wang, K.; Sun, J.; Qiu, Y.
2013-07-01
Monte Carlo simulations containing a large number of tallies generally suffer severe performance penalties due to a significant amount of run time spent in searching for and scoring individual tally bins. This paper describes the improved methods of handling large numbers of tallies, which have been implemented in the RMC Monte Carlo code. The calculation results demonstrate that the proposed methods can considerably improve the tally performance when massive tallies are treated. In the calculated case with 6 million of tally regions, only 10% of run time is increased in each active cycle against each inactive cycle. (authors)
Development of a Monte-Carlo Radiative Transfer Code for the Juno/JIRAM Limb Measurements
NASA Astrophysics Data System (ADS)
Sindoni, G.; Adriani, A.; Mayorov, B.; Aoki, S.; Grassi, D.; Moriconi, M.; Oliva, F.
2013-09-01
The Juno/JIRAM instrument will acquire limb spectra of the Jupiter atmosphere in the infrared spectral range. The analysis of these spectra requires a radiative transfer code that takes into account the multiple scattering by particles in a spherical-shell atmosphere. Therefore, we are developing a code based on the Monte-Carlo approach to simulate the JIRAM observations. The validation of the code was performed by comparison with DISORT-based codes.
Matthew Ellis; Derek Gaston; Benoit Forget; Kord Smith
2011-07-01
In recent years the use of Monte Carlo methods for modeling reactors has become feasible due to the increasing availability of massively parallel computer systems. One of the primary challenges yet to be fully resolved, however, is the efficient and accurate inclusion of multiphysics feedback in Monte Carlo simulations. The research in this paper presents a preliminary coupling of the open source Monte Carlo code OpenMC with the open source Multiphysics Object-Oriented Simulation Environment (MOOSE). The coupling of OpenMC and MOOSE will be used to investigate efficient and accurate numerical methods needed to include multiphysics feedback in Monte Carlo codes. An investigation into the sensitivity of Doppler feedback to fuel temperature approximations using a two dimensional 17x17 PWR fuel assembly is presented in this paper. The results show a functioning multiphysics coupling between OpenMC and MOOSE. The coupling utilizes Functional Expansion Tallies to accurately and efficiently transfer pin power distributions tallied in OpenMC to unstructured finite element meshes used in MOOSE. The two dimensional PWR fuel assembly case also demonstrates that for a simplified model the pin-by-pin doppler feedback can be adequately replicated by scaling a representative pin based on pin relative powers.
The analog linear interpolation approach for Monte Carlo simulation of PGNAA: The CEARPGA code
NASA Astrophysics Data System (ADS)
Zhang, Wenchao; Gardner, Robin P.
2004-01-01
The analog linear interpolation approach (ALI) has been developed and implemented to eliminate the big weight problem in the Monte Carlo simulation code CEARPGA. The CEARPGA code was previously developed to generate elemental library spectra for using the Monte Carlo - library least-squares (MCLLS) approach in prompt gamma-ray neutron activation analysis (PGNAA). In addition, some other improvements to this code have been introduced, including (1) adopting the latest photon cross-section data, (2) using an improved detector response function, (3) adding the neutron activation backgrounds, (4) generating the individual natural background libraries, (5) adding the tracking of annihilation photons from pair production interactions outside of the detector and (6) adopting a general geometry package. The simulated result from the new CEARPGA code is compared with those calculated from the previous CEARPGA code and the MCNP code and experimental data. The new CEARPGA code is found to give the best result.
TART97 a coupled neutron-photon 3-D, combinatorial geometry Monte Carlo transport code
Cullen, D.E.
1997-11-22
TART97 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo transport code. This code can on any modern computer. It is a complete system to assist you with input preparation, running Monte Carlo calculations, and analysis of output results. TART97 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART97 is distributed on CD. This CD contains on- line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART97 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART97 and its data riles.
Implementation of a Monte Carlo based inverse planning model for clinical IMRT with MCNP code
NASA Astrophysics Data System (ADS)
He, Tongming Tony
In IMRT inverse planning, inaccurate dose calculations and limitations in optimization algorithms introduce both systematic and convergence errors to treatment plans. The goal of this work is to practically implement a Monte Carlo based inverse planning model for clinical IMRT. The intention is to minimize both types of error in inverse planning and obtain treatment plans with better clinical accuracy than non-Monte Carlo based systems. The strategy is to calculate the dose matrices of small beamlets by using a Monte Carlo based method. Optimization of beamlet intensities is followed based on the calculated dose data using an optimization algorithm that is capable of escape from local minima and prevents possible pre-mature convergence. The MCNP 4B Monte Carlo code is improved to perform fast particle transport and dose tallying in lattice cells by adopting a selective transport and tallying algorithm. Efficient dose matrix calculation for small beamlets is made possible by adopting a scheme that allows concurrent calculation of multiple beamlets of single port. A finite-sized point source (FSPS) beam model is introduced for easy and accurate beam modeling. A DVH based objective function and a parallel platform based algorithm are developed for the optimization of intensities. The calculation accuracy of improved MCNP code and FSPS beam model is validated by dose measurements in phantoms. Agreements better than 1.5% or 0.2 cm have been achieved. Applications of the implemented model to clinical cases of brain, head/neck, lung, spine, pancreas and prostate have demonstrated the feasibility and capability of Monte Carlo based inverse planning for clinical IMRT. Dose distributions of selected treatment plans from a commercial non-Monte Carlo based system are evaluated in comparison with Monte Carlo based calculations. Systematic errors of up to 12% in tumor doses and up to 17% in critical structure doses have been observed. The clinical importance of Monte Carlo based
Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System
Karim, Julia Abdul
2008-05-20
The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained.
Large scale cratering of the lunar highlands - Some Monte Carlo model considerations
NASA Technical Reports Server (NTRS)
Hoerz, F.; Gibbons, R. V.; Hill, R. E.; Gault, D. E.
1976-01-01
In an attempt to understand the scale and intensity of the moon's early, large scale meteoritic bombardment, a Monte Carlo computer model simulated the effects of all lunar craters greater than 800 m in diameter, for example, the number of times and depths specific fractions of the entire lunar surface were cratered. The model used observed crater size frequencies and crater-geometries compatible with the suggestions of Pike (1974) and Dence (1973); it simulated bombardment histories up to a factor of 10 more intense than those reflected by the present-day crater number density of the lunar highlands. For the present-day cratering record the model yields the following: approximately 25% of the entire lunar surface has not been cratered deeper than 100 m; 50% may have been cratered to 2-3 km depth; less than 5% of the surface has been cratered deeper than about 15 km. A typical highland site has suffered 1-2 impacts. Corresponding values for more intense bombardment histories are also presented, though it must remain uncertain what the absolute intensity of the moon's early meteorite bombardment was.
Progress and status of the OpenMC Monte Carlo code
Romano, P. K.; Herman, B. R.; Horelik, N. E.; Forget, B.; Smith, K.; Siegel, A. R.
2013-07-01
The present work describes the latest advances and progress in the development of the OpenMC Monte Carlo code, an open-source code originating from the Massachusetts Institute of Technology. First, an overview of the development workflow of OpenMC is given. Various enhancements to the code such as real-time XML input validation, state points, plotting, OpenMP threading, and coarse mesh finite difference acceleration are described. (authors)
A General-Purpose Monte Carlo Gamma-Ray Transport Code System for Minicomputers.
Energy Science and Technology Software Center (ESTSC)
1981-08-27
Version 00 The OGRE code system was designed to calculate, by Monte Carlo methods, any quantity related to gamma-ray transport. The system is represented by two codes which treat slab geometry. OGRE-P1 computes the dose on one side of a slab for a source on the other side, and HOTONE computes energy deposition in addition. The source may be monodirectional, isotropic, or cosine distributed.
Perfetti, Christopher M; Martin, William R; Rearden, Bradley T; Williams, Mark L
2012-01-01
Three methods for calculating continuous-energy eigenvalue sensitivity coefficients were developed and implemented into the SHIFT Monte Carlo code within the Scale code package. The methods were used for several simple test problems and were evaluated in terms of speed, accuracy, efficiency, and memory requirements. A promising new method for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was developed and produced accurate sensitivity coefficients with figures of merit that were several orders of magnitude larger than those from existing methods.
Dose conversion coefficients for ICRP110 voxel phantom in the Geant4 Monte Carlo code
NASA Astrophysics Data System (ADS)
Martins, M. C.; Cordeiro, T. P. V.; Silva, A. X.; Souza-Santos, D.; Queiroz-Filho, P. P.; Hunt, J. G.
2014-02-01
The reference adult male voxel phantom recommended by International Commission on Radiological Protection no. 110 was implemented in the Geant4 Monte Carlo code. Geant4 was used to calculate Dose Conversion Coefficients (DCCs) expressed as dose deposited in organs per air kerma for photons, electrons and neutrons in the Annals of the ICRP. In this work the AP and PA irradiation geometries of the ICRP male phantom were simulated for the purpose of benchmarking the Geant4 code. Monoenergetic photons were simulated between 15 keV and 10 MeV and the results were compared with ICRP 110, the VMC Monte Carlo code and the literature data available, presenting a good agreement.
Monte Carlo Code System for Electron (Positron) Dose Kernel Calculations.
CHIBANI, OMAR
1999-05-12
Version 00 KERNEL performs dose kernel calculations for an electron (positron) isotropic point source in an infinite homogeneous medium. First, the auxiliary code PRELIM is used to prepare cross section data for the considered medium. Then the KERNEL code simulates the transport of electrons and bremsstrahlung photons through the medium until all particles reach their cutoff energies. The deposited energy is scored in concentric spherical shells at a radial distance ranging from zero to twice the source particle range.
MCNP: a general Monte Carlo code for neutron and photon transport
Forster, R.A.; Godfrey, T.N.K.
1985-01-01
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.
Update on the Development and Validation of MERCURY: A Modern, Monte Carlo Particle Transport Code
Procassini, R J; Taylor, J M; McKinley, M S; Greenman, G M; Cullen, D E; O'Brien, M J; Beck, B R; Hagmann, C A
2005-06-06
An update on the development and validation of the MERCURY Monte Carlo particle transport code is presented. MERCURY is a modern, parallel, general-purpose Monte Carlo code being developed at the Lawrence Livermore National Laboratory. During the past year, several major algorithm enhancements have been completed. These include the addition of particle trackers for 3-D combinatorial geometry (CG), 1-D radial meshes, 2-D quadrilateral unstructured meshes, as well as a feature known as templates for defining recursive, repeated structures in CG. New physics capabilities include an elastic-scattering neutron thermalization model, support for continuous energy cross sections and S ({alpha}, {beta}) molecular bound scattering. Each of these new physics features has been validated through code-to-code comparisons with another Monte Carlo transport code. Several important computer science features have been developed, including an extensible input-parameter parser based upon the XML data description language, and a dynamic load-balance methodology for efficient parallel calculations. This paper discusses the recent work in each of these areas, and describes a plan for future extensions that are required to meet the needs of our ever expanding user base.
Monte Carlo Code System for Electron (Positron) Dose Kernel Calculations.
Energy Science and Technology Software Center (ESTSC)
1999-05-12
Version 00 KERNEL performs dose kernel calculations for an electron (positron) isotropic point source in an infinite homogeneous medium. First, the auxiliary code PRELIM is used to prepare cross section data for the considered medium. Then the KERNEL code simulates the transport of electrons and bremsstrahlung photons through the medium until all particles reach their cutoff energies. The deposited energy is scored in concentric spherical shells at a radial distance ranging from zero to twicemore » the source particle range.« less
PEREGRINE: An all-particle Monte Carlo code for radiation therapy
Hartmann Siantar, C.L.; Chandler, W.P.; Rathkopf, J.A.; Svatos, M.M.; White, R.M.
1994-09-01
The goal of radiation therapy is to deliver a lethal dose to the tumor while minimizing the dose to normal tissues. To carry out this task, it is critical to calculate correctly the distribution of dose delivered. Monte Carlo transport methods have the potential to provide more accurate prediction of dose distributions than currently-used methods. PEREGRINE is a new Monte Carlo transport code developed at Lawrence Livermore National Laboratory for the specific purpose of modeling the effects of radiation therapy. PEREGRINE transports neutrons, photons, electrons, positrons, and heavy charged-particles, including protons, deuterons, tritons, helium-3, and alpha particles. This paper describes the PEREGRINE transport code and some preliminary results for clinically relevant materials and radiation sources.
Buck, R M; Hall, J M
1999-06-01
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.
NASA Technical Reports Server (NTRS)
Platt, M. E.; Lewis, E. E.; Boehm, F.
1991-01-01
A Monte Carlo Fortran computer program was developed that uses two variance reduction techniques for computing system reliability applicable to solving very large highly reliable fault-tolerant systems. The program is consistent with the hybrid automated reliability predictor (HARP) code which employs behavioral decomposition and complex fault-error handling models. This new capability is called MC-HARP which efficiently solves reliability models with non-constant failures rates (Weibull). Common mode failure modeling is also a specialty.
The Serpent Monte Carlo Code: Status, Development and Applications in 2013
NASA Astrophysics Data System (ADS)
Leppänen, Jaakko; Pusa, Maria; Viitanen, Tuomas; Valtavirta, Ville; Kaltiaisenaho, Toni
2014-06-01
The Serpent Monte Carlo reactor physics burnup calculation code has been developed at VTT Technical Research Centre of Finland since 2004, and is currently used in 100 universities and research organizations around the world. This paper presents the brief history of the project, together with the currently available methods and capabilities and plans for future work. Typical user applications are introduced in the form of a summary review on Serpent-related publications over the past few years.
Brown, F.B.; Sutton, T.M.
1996-02-01
This report is composed of the lecture notes from the first half of a 32-hour graduate-level course on Monte Carlo methods offered at KAPL. These notes, prepared by two of the principle developers of KAPL`s RACER Monte Carlo code, cover the fundamental theory, concepts, and practices for Monte Carlo analysis. In particular, a thorough grounding in the basic fundamentals of Monte Carlo methods is presented, including random number generation, random sampling, the Monte Carlo approach to solving transport problems, computational geometry, collision physics, tallies, and eigenvalue calculations. Furthermore, modern computational algorithms for vector and parallel approaches to Monte Carlo calculations are covered in detail, including fundamental parallel and vector concepts, the event-based algorithm, master/slave schemes, parallel scaling laws, and portability issues.
Monte Carlo code for neutron scattering instrumentation design and analysis
Daemen, L.; Fitzsimmons, M.; Hjelm, R.; Olah, G.; Roberts, J.; Seeger, P.; Smith, G.; Thelliez, T.
1996-09-01
This is the final report of a one-year, Laboratory-Directed Research and Development (LDRD) at the Los Alamos National Laboratory (LANL). The development of next generation, accelerator based neutron sources calls for the design of new instruments for neutron scattering studies of materials. It will be necessary, in the near future, to evaluate accurately and rapidly the performance of new and traditional neutron instruments at short- and long-pulse spallation neutron sources, as well as continuous sources. We have developed a code that is a design tool to assist the instrument designer model new or existing instruments, test their performance, and optimize their most important features.
Improvements of MCOR: A Monte Carlo depletion code system for fuel assembly reference calculations
Tippayakul, C.; Ivanov, K.; Misu, S.
2006-07-01
This paper presents the improvements of MCOR, a Monte Carlo depletion code system for fuel assembly reference calculations. The improvements of MCOR were initiated by the cooperation between the Penn State Univ. and AREVA NP to enhance the original Penn State Univ. MCOR version in order to be used as a new Monte Carlo depletion analysis tool. Essentially, a new depletion module using KORIGEN is utilized to replace the existing ORIGEN-S depletion module in MCOR. Furthermore, the online burnup cross section generation by the Monte Carlo calculation is implemented in the improved version instead of using the burnup cross section library pre-generated by a transport code. Other code features have also been added to make the new MCOR version easier to use. This paper, in addition, presents the result comparisons of the original and the improved MCOR versions against CASMO-4 and OCTOPUS. It was observed in the comparisons that there were quite significant improvements of the results in terms of k{sub inf}, fission rate distributions and isotopic contents. (authors)
A user's manual for MASH 1. 0: A Monte Carlo Adjoint Shielding Code System
Johnson, J.O.
1992-03-01
The Monte Carlo Adjoint Shielding Code System, MASH, calculates neutron and gamma-ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air-over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system include the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. MASH is the successor to the Vehicle Code System (VCS) initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the dose importance'' of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response a a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem (input data and selected output edits) for each code.
Pandya, Tara M.; Johnson, Seth R.; Evans, Thomas M.; Davidson, Gregory G.; Hamilton, Steven P.; Godfrey, Andrew T.
2015-12-21
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. Somemore » 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.« less
Pandya, Tara M.; Johnson, Seth R.; Evans, Thomas M.; Davidson, Gregory G.; Hamilton, Steven P.; Godfrey, Andrew T.
2015-12-21
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.
An object-oriented implementation of a parallel Monte Carlo code for radiation transport
NASA Astrophysics Data System (ADS)
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.
NASA Astrophysics Data System (ADS)
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.
Li, Junli; Li, Chunyan; Qiu, Rui; Yan, Congchong; Xie, Wenzhang; Wu, Zhen; Zeng, Zhi; Tung, Chuanjong
2015-09-01
The method of Monte Carlo simulation is a powerful tool to investigate the details of radiation biological damage at the molecular level. In this paper, a Monte Carlo code called NASIC (Nanodosimetry Monte Carlo Simulation Code) was developed. It includes physical module, pre-chemical module, chemical module, geometric module and DNA damage module. The physical module can simulate physical tracks of low-energy electrons in the liquid water event-by-event. More than one set of inelastic cross sections were calculated by applying the dielectric function method of Emfietzoglou's optical-data treatments, with different optical data sets and dispersion models. In the pre-chemical module, the ionised and excited water molecules undergo dissociation processes. In the chemical module, the produced radiolytic chemical species diffuse and react. In the geometric module, an atomic model of 46 chromatin fibres in a spherical nucleus of human lymphocyte was established. In the DNA damage module, the direct damages induced by the energy depositions of the electrons and the indirect damages induced by the radiolytic chemical species were calculated. The parameters should be adjusted to make the simulation results be agreed with the experimental results. In this paper, the influence study of the inelastic cross sections and vibrational excitation reaction on the parameters and the DNA strand break yields were studied. Further work of NASIC is underway. PMID:25883312
Monte Carlo simulation of MOSFET dosimeter for electron backscatter using the GEANT4 code.
Chow, James C L; Leung, Michael K K
2008-06-01
The aim of this study is to investigate the influence of the body of the metal-oxide-semiconductor field effect transistor (MOSFET) dosimeter in measuring the electron backscatter from lead. The electron backscatter factor (EBF), which is defined as the ratio of dose at the tissue-lead interface to the dose at the same point without the presence of backscatter, was calculated by the Monte Carlo simulation using the GEANT4 code. Electron beams with energies of 4, 6, 9, and 12 MeV were used in the simulation. It was found that in the presence of the MOSFET body, the EBFs were underestimated by about 2%-0.9% for electron beam energies of 4-12 MeV, respectively. The trend of the decrease of EBF with an increase of electron energy can be explained by the small MOSFET dosimeter, mainly made of epoxy and silicon, not only attenuated the electron fluence of the electron beam from upstream, but also the electron backscatter generated by the lead underneath the dosimeter. However, this variation of the EBF underestimation is within the same order of the statistical uncertainties as the Monte Carlo simulations, which ranged from 1.3% to 0.8% for the electron energies of 4-12 MeV, due to the small dosimetric volume. Such small EBF deviation is therefore insignificant when the uncertainty of the Monte Carlo simulation is taken into account. Corresponding measurements were carried out and uncertainties compared to Monte Carlo results were within +/- 2%. Spectra of energy deposited by the backscattered electrons in dosimetric volumes with and without the lead and MOSFET were determined by Monte Carlo simulations. It was found that in both cases, when the MOSFET body is either present or absent in the simulation, deviations of electron energy spectra with and without the lead decrease with an increase of the electron beam energy. Moreover, the softer spectrum of the backscattered electron when lead is present can result in a reduction of the MOSFET response due to stronger
A User's Manual for MASH V1.5 - A Monte Carlo Adjoint Shielding Code System
C. O. Slater; J. M. Barnes; J. O. Johnson; J.D. Drischler
1998-10-01
The Monte Carlo ~djoint ~ielding Code System, MASH, calculates neutron and gamma- ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air- over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system includes the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. The current version, MASH v 1.5, is the successor to the original MASH v 1.0 code system initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the "dose importance" of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response as a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem.
Movable geometry and eigenvalue search capability in the MC21 Monte Carlo code
Gill, D. F.; Nease, B. R.; Griesheimer, D. P.
2013-07-01
A description of a robust and flexible movable geometry implementation in the Monte Carlo code MC21 is described along with a search algorithm that can be used in conjunction with the movable geometry capability to perform eigenvalue searches based on the position of some geometric component. The natural use of the combined movement and search capability is searching to critical through variation of control rod (or control drum) position. The movable geometry discussion provides the mathematical framework for moving surfaces in the MC21 combinatorial solid geometry description. A discussion of the interface between the movable geometry system and the user is also described, particularly the ability to create a hierarchy of movable groups. Combined with the hierarchical geometry description in MC21 the movable group framework provides a very powerful system for inline geometry modification. The eigenvalue search algorithm implemented in MC21 is also described. The foundations of this algorithm are a regula falsi search though several considerations are made in an effort to increase the efficiency of the algorithm for use with Monte Carlo. Specifically, criteria are developed to determine after each batch whether the Monte Carlo calculation should be continued, the search iteration can be rejected, or the search iteration has converged. These criteria seek to minimize the amount of time spent per iteration. Results for the regula falsi method are shown, illustrating that the method as implemented is indeed convergent and that the optimizations made ultimately reduce the total computational expense. (authors)
Hart, S. W. D.; Maldonado, G. Ivan; Celik, Cihangir; Leal, Luiz C
2014-01-01
For many Monte Carlo codes cross sections are generally only created at a set of predetermined temperatures. This causes an increase in error as one moves further and further away from these temperatures in the Monte Carlo model. This paper discusses recent progress in the Scale Monte Carlo module KENO to create problem dependent, Doppler broadened, cross sections. Currently only broadening the 1D cross sections and probability tables is addressed. The approach uses a finite difference method to calculate the temperature dependent cross-sections for the 1D data, and a simple linear-logarithmic interpolation in the square root of temperature for the probability tables. Work is also ongoing to address broadening theS (alpha , beta) tables. With the current approach the temperature dependent cross sections are Doppler broadened before transport starts, and, for all but a few isotopes, the impact on cross section loading is negligible. Results can be compared with those obtained by using multigroup libraries, as KENO currently does interpolation on the multigroup cross sections to determine temperature dependent cross-sections. Current results compare favorably with these expected results.
A Monte Carlo model for the gardening of the lunar regolith
NASA Technical Reports Server (NTRS)
Arnold, J. R.
1975-01-01
The processes of movement and turnover of the lunar regolith are described by a Monte Carlo model. The movement of material by the direct cratering process is the dominant mode, but slumping is also included for angles exceeding the static angle of repose. Using a group of interrelated computer programs, a large number of properties are calculated, including topography, formation of layers, depth of the disturbed layer, nuclear-track distributions, and cosmogenic nuclides. In the most complex program, the history of a 36-point square array is followed for times up to 400 million years. The histories generated are complex and exhibit great variety. Because a crater covers much less area than its ejecta blanket, there is a tendency for the height change at a test point to exhibit periods of slow accumulation followed by sudden excavation. In general, the agreement with experiment and observation seems good, but two areas of disagreement stand out. First, the calculated surface is rougher than that observed. Second, the observed bombardment ages, of the order 400 million are shorter than expected (by perhaps a factor of 5).
Perfetti, C.; Martin, W.; Rearden, B.; Williams, M.
2012-07-01
Three methods for calculating continuous-energy eigenvalue sensitivity coefficients were developed and implemented into the Shift Monte Carlo code within the SCALE code package. The methods were used for two small-scale test problems and were evaluated in terms of speed, accuracy, efficiency, and memory requirements. A promising new method for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was developed and produced accurate sensitivity coefficients with figures of merit that were several orders of magnitude larger than those from existing methods. (authors)
A Coupled Neutron-Photon 3-D Combinatorial Geometry Monte Carlo Transport Code
Energy Science and Technology Software Center (ESTSC)
1998-06-12
TART97 is a coupled neutron-photon, 3 dimensional, combinatorial geometry, time dependent Monte Carlo 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. TART97 is also incredibly fast: if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system canmore » save you a great deal of time and energy. TART 97 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART97 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART97 and ist data files.« less
Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes
Smith, L.M.; Hochstedler, R.D.
1997-02-01
Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code).
A fast Monte Carlo code for proton transport in radiation therapy based on MCNPX
Jabbari, Keyvan; Seuntjens, Jan
2014-01-01
An important requirement for proton therapy is a software for dose calculation. Monte Carlo is the most accurate method for dose calculation, but it is very slow. In this work, a method is developed to improve the speed of dose calculation. The method is based on pre-generated tracks for particle transport. The MCNPX code has been used for generation of tracks. A set of data including the track of the particle was produced in each particular material (water, air, lung tissue, bone, and soft tissue). This code can transport protons in wide range of energies (up to 200 MeV for proton). The validity of the fast Monte Carlo (MC) code is evaluated with data MCNPX as a reference code. While analytical pencil beam algorithm transport shows great errors (up to 10%) near small high density heterogeneities, there was less than 2% deviation of MCNPX results in our dose calculation and isodose distribution. In terms of speed, the code runs 200 times faster than MCNPX. In the Fast MC code which is developed in this work, it takes the system less than 2 minutes to calculate dose for 106 particles in an Intel Core 2 Duo 2.66 GHZ desktop computer. PMID:25190994
Application of S{sub N} and Monte Carlo codes to the SHEBA critical assemblies
O`Dell, R.D.
1993-07-01
The Solution High-Energy Burst Assembly (SHEBA) at Los Alamos is a low-enriched (4.95 wt. %) aqueous uranyl fluoride solution critical assembly. There are two SHEBA configurations, both consisting of right circular cylinders with a central control rod. The first configuration, hereafter called the old SHEBA, had a fuel solution diameter of 54.6 cm and a measured critical solution height of 36.5 cm. An improved modification, hereafter called the new SHEBA, has a fuel solution diameter of 48.9 cm but since it is not yet operational, the critical solution height has not yet been measured. In this presentation the application of the discrete-ordinates (S{sub N}) code TWODANT using Hansen-Roach cross sections and the MCNP Monte Carlo code using continuous-energy cross sections for calculating the critical solution heights for both the old and new SHEBA assemblies is described. The code`s predictions are compared and it is shown that a single calculation with a standard computer code may yield misleading results, especially when using a Monte Carlo code.
OpenMC: A State-of-the-Art Monte Carlo Code for Research and Development
NASA Astrophysics Data System (ADS)
Romano, Paul K.; Horelik, Nicholas E.; Herman, Bryan R.; Nelson, Adam G.; Forget, Benoit; Smith, Kord
2014-06-01
This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently developed at the Massachusetts Institute of Technology. OpenMC uses continuous-energy cross sections and a constructive solid geometry representation, enabling high-fidelity modeling of nuclear reactors and other systems. Modern, portable input/output file formats are used in OpenMC: XML for input, and HDF5 for output. High performance parallel algorithms in OpenMC have demonstrated near-linear scaling to over 100,000 processors on modern supercomputers. Other topics discussed in this paper include plotting, CMFD acceleration, variance reduction, eigenvalue calculations, and software development processes.
Development of a dynamic simulation mode in Serpent 2 Monte Carlo code
Leppaenen, J.
2013-07-01
This paper presents a dynamic neutron transport mode, currently being implemented in the Serpent 2 Monte Carlo code for the purpose of simulating short reactivity transients with temperature feedback. The transport routine is introduced and validated by comparison to MCNP5 calculations. The method is also tested in combination with an internal temperature feedback module, which forms the inner part of a multi-physics coupling scheme in Serpent 2. The demo case for the coupled calculation is a reactivity-initiated accident (RIA) in PWR fuel. (authors)
New Capabilities in Mercury: A Modern, Monte Carlo Particle Transport Code
Procassini, R J; Cullen, D E; Greenman, G M; Hagmann, C A; Kramer, K J; McKinley, M S; O'Brien, M J; Taylor, J M
2007-03-08
The new physics, algorithmic and computer science capabilities of the Mercury general-purpose Monte Carlo particle transport code are discussed. The new physics and algorithmic features include in-line energy deposition and isotopic depletion, significant enhancements to the tally and source capabilities, diagnostic ray-traced particles, support for multi-region hybrid (mesh and combinatorial geometry) systems, and a probability of initiation method. Computer science enhancements include a second method of dynamically load-balancing parallel calculations, improved methods for visualizing 3-D combinatorial geometries and initial implementation of an in-line visualization capabilities.
Development and validation of ALEPH2 Monte Carlo burn-up code
Van Den Eynde, G.; Stankovskiy, A.; Fiorito, L.; Broustaut, M.
2013-07-01
The ALEPH2 Monte Carlo depletion code has two principal features that make it a flexible and powerful tool for reactor analysis. First of all, its comprehensive nuclear data library ensures the consistency between steady-state Monte Carlo and deterministic depletion modules. It covers neutron and proton induced reactions, neutron and proton fission product yields, spontaneous fission product yields, radioactive decay data and total recoverable energies per fission. Secondly, ALEPH2 uses an advanced numerical solver for the first order ordinary differential equations describing the isotope balances, namely a Radau IIA implicit Runge-Kutta method. The versatility of the code allows using it for time behavior simulation of various systems ranging from single pin model to full-scale reactor model. The code is extensively used for the neutronics design of the MYRRHA research fast spectrum facility which will operate in both critical and sub-critical modes. The code has been validated on the decay heat data from JOYO experimental fast reactor. (authors)
Monte Carlo N Particle code - Dose distribution of clinical electron beams in inhomogeneous phantoms
Nedaie, H. A.; Mosleh-Shirazi, M. A.; Allahverdi, M.
2013-01-01
Electron dose distributions calculated using the currently available analytical methods can be associated with large uncertainties. The Monte Carlo method is the most accurate method for dose calculation in electron beams. Most of the clinical electron beam simulation studies have been performed using non- MCNP [Monte Carlo N Particle] codes. Given the differences between Monte Carlo codes, this work aims to evaluate the accuracy of MCNP4C-simulated electron dose distributions in a homogenous phantom and around inhomogeneities. Different types of phantoms ranging in complexity were used; namely, a homogeneous water phantom and phantoms made of polymethyl methacrylate slabs containing different-sized, low- and high-density inserts of heterogeneous materials. Electron beams with 8 and 15 MeV nominal energy generated by an Elekta Synergy linear accelerator were investigated. Measurements were performed for a 10 cm × 10 cm applicator at a source-to-surface distance of 100 cm. Individual parts of the beam-defining system were introduced into the simulation one at a time in order to show their effect on depth doses. In contrast to the first scattering foil, the secondary scattering foil, X and Y jaws and applicator provide up to 5% of the dose. A 2%/2 mm agreement between MCNP and measurements was found in the homogenous phantom, and in the presence of heterogeneities in the range of 1-3%, being generally within 2% of the measurements for both energies in a "complex" phantom. A full-component simulation is necessary in order to obtain a realistic model of the beam. The MCNP4C results agree well with the measured electron dose distributions. PMID:23533162
Nedaie, H A; Mosleh-Shirazi, M A; Allahverdi, M
2013-01-01
Electron dose distributions calculated using the currently available analytical methods can be associated with large uncertainties. The Monte Carlo method is the most accurate method for dose calculation in electron beams. Most of the clinical electron beam simulation studies have been performed using non- MCNP [Monte Carlo N Particle] codes. Given the differences between Monte Carlo codes, this work aims to evaluate the accuracy of MCNP4C-simulated electron dose distributions in a homogenous phantom and around inhomogeneities. Different types of phantoms ranging in complexity were used; namely, a homogeneous water phantom and phantoms made of polymethyl methacrylate slabs containing different-sized, low- and high-density inserts of heterogeneous materials. Electron beams with 8 and 15 MeV nominal energy generated by an Elekta Synergy linear accelerator were investigated. Measurements were performed for a 10 cm × 10 cm applicator at a source-to-surface distance of 100 cm. Individual parts of the beam-defining system were introduced into the simulation one at a time in order to show their effect on depth doses. In contrast to the first scattering foil, the secondary scattering foil, X and Y jaws and applicator provide up to 5% of the dose. A 2%/2 mm agreement between MCNP and measurements was found in the homogenous phantom, and in the presence of heterogeneities in the range of 1-3%, being generally within 2% of the measurements for both energies in a "complex" phantom. A full-component simulation is necessary in order to obtain a realistic model of the beam. The MCNP4C results agree well with the measured electron dose distributions. PMID:23533162
Applications of FLUKA Monte Carlo Code for Nuclear and Accelerator Physics
Battistoni, Giuseppe; Broggi, Francesco; Brugger, Markus; Campanella, Mauro; Carboni, Massimo; Empl, Anton; Fasso, Alberto; Gadioli, Ettore; Cerutti, Francesco; Ferrari, Alfredo; Ferrari, Anna; Lantz, Matthias; Mairani, Andrea; Margiotta, M.; Morone, Christina; Muraro, Silvia; Parodi, Katerina; Patera, Vincenzo; Pelliccioni, Maurizio; Pinsky, Lawrence; Ranft, Johannes; /Siegen U. /CERN /Seibersdorf, Reaktorzentrum /INFN, Milan /Milan U. /SLAC /INFN, Legnaro /INFN, Bologna /Bologna U. /CERN /HITS, Heidelberg /CERN /CERN /Frascati /CERN /CERN /CERN /CERN /NASA, Houston
2012-04-17
FLUKA is a general purpose Monte Carlo code capable of handling all radiation components from thermal energies (for neutrons) or 1 keV (for all other particles) to cosmic ray energies and can be applied in many different fields. Presently the code is maintained on Linux. The validity of the physical models implemented in FLUKA has been benchmarked against a variety of experimental data over a wide energy range, from accelerator data to cosmic ray showers in the Earth atmosphere. FLUKA is widely used for studies related both to basic research and to applications in particle accelerators, radiation protection and dosimetry, including the specific issue of radiation damage in space missions, radiobiology (including radiotherapy) and cosmic ray calculations. After a short description of the main features that make FLUKA valuable for these topics, the present paper summarizes some of the recent applications of the FLUKA Monte Carlo code in the nuclear as well high energy physics. In particular it addresses such topics as accelerator related applications.
Applications of FLUKA Monte Carlo code for nuclear and accelerator physics
NASA Astrophysics Data System (ADS)
Battistoni, Giuseppe; Broggi, Francesco; Brugger, Markus; Campanella, Mauro; Carboni, Massimo; Empl, Anton; Fassò, Alberto; Gadioli, Ettore; Cerutti, Francesco; Ferrari, Alfredo; Ferrari, Anna; Lantz, Matthias; Mairani, Andrea; Margiotta, M.; Morone, Cristina; Muraro, Silvia; Parodi, Katia; Patera, Vincenzo; Pelliccioni, Mauricio; Pinsky, Larry; Ranft, Johannes; Roesler, Stefan; Rollet, Sofia; Sala, Paola R.; Santana, Mario; Sarchiapone, Lucia; Sioli, Massimiliano; Smirnov, George; Sommerer, Florian; Theis, Christian; Trovati, Stefania; Villari, R.; Vincke, Heinz; Vincke, Helmut; Vlachoudis, Vasilis; Vollaire, Joachim; Zapp, Neil
2011-12-01
FLUKA is a general purpose Monte Carlo code capable of handling all radiation components from thermal energies (for neutrons) or 1 keV (for all other particles) to cosmic ray energies and can be applied in many different fields. Presently the code is maintained on Linux. The validity of the physical models implemented in FLUKA has been benchmarked against a variety of experimental data over a wide energy range, from accelerator data to cosmic ray showers in the Earth atmosphere. FLUKA is widely used for studies related both to basic research and to applications in particle accelerators, radiation protection and dosimetry, including the specific issue of radiation damage in space missions, radiobiology (including radiotherapy) and cosmic ray calculations. After a short description of the main features that make FLUKA valuable for these topics, the present paper summarizes some of the recent applications of the FLUKA Monte Carlo code in the nuclear as well high energy physics. In particular it addresses such topics as accelerator related applications.
Monte Carlo Code System for High-Energy Radiation Transport Calculations.
Energy Science and Technology Software Center (ESTSC)
2000-02-16
Version 00 HERMES-KFA consists of a set of Monte Carlo Codes used to simulate particle radiation and interaction with matter. The main codes are HETC, MORSE, and EGS. They are supported by a common geometry package, common random routines, a command interpreter, and auxiliary codes like NDEM that is used to generate a gamma-ray source from nuclear de-excitation after spallation processes. The codes have been modified so that any particle history falling outside the domainmore » of the physical theory of one program can be submitted to another program in the suite to complete the work. Also response data can be submitted by each program, to be collected and combined by a statistic package included within the command interpreter.« less
NASA Astrophysics Data System (ADS)
Chauvet, Yves
1985-07-01
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.
Nuclear data processing for energy release and deposition calculations in the MC21 Monte Carlo code
Trumbull, T. H.
2013-07-01
With the recent emphasis in performing multiphysics calculations using Monte Carlo transport codes such as MC21, the need for accurate estimates of the energy deposition-and the subsequent heating - has increased. However, the availability and quality of data necessary to enable accurate neutron and photon energy deposition calculations can be an issue. A comprehensive method for handling the nuclear data required for energy deposition calculations in MC21 has been developed using the NDEX nuclear data processing system and leveraging the capabilities of NJOY. The method provides a collection of data to the MC21 Monte Carlo code supporting the computation of a wide variety of energy release and deposition tallies while also allowing calculations with different levels of fidelity to be performed. Detailed discussions on the usage of the various components of the energy release data are provided to demonstrate novel methods in borrowing photon production data, correcting for negative energy release quantities, and adjusting Q values when necessary to preserve energy balance. Since energy deposition within a reactor is a result of both neutron and photon interactions with materials, a discussion on the photon energy deposition data processing is also provided. (authors)
Spread-out Bragg peak and monitor units calculation with the Monte Carlo Code MCNPX
Herault, J.; Iborra, N.; Serrano, B.; Chauvel, P.
2007-02-15
The aim of this work was to study the dosimetric potential of the Monte Carlo code MCNPX applied to the protontherapy field. For series of clinical configurations a comparison between simulated and experimental data was carried out, using the proton beam line of the MEDICYC isochronous cyclotron installed in the Centre Antoine Lacassagne in Nice. The dosimetric quantities tested were depth-dose distributions, output factors, and monitor units. For each parameter, the simulation reproduced accurately the experiment, which attests the quality of the choices made both in the geometrical description and in the physics parameters for beam definition. These encouraging results enable us today to consider a simplification of quality control measurements in the future. Monitor Units calculation is planned to be carried out with preestablished Monte Carlo simulation data. The measurement, which was until now our main patient dose calibration system, will be progressively replaced by computation based on the MCNPX code. This determination of Monitor Units will be controlled by an independent semi-empirical calculation.
Bianchini, G.; Burgio, N.; Carta, M.; Peluso, V.; Fabrizio, V.; Ricci, L.
2012-07-01
The GUINEVERE experiment (Generation of Uninterrupted Intense Neutrons at the lead Venus Reactor) is an experimental program in support of the ADS technology presently carried out at SCK-CEN in Mol (Belgium). In the experiment a modified lay-out of the original thermal VENUS critical facility is coupled to an accelerator, built by the French body CNRS in Grenoble, working in both continuous and pulsed mode and delivering 14 MeV neutrons by bombardment of deuterons on a tritium-target. The modified lay-out of the facility consists of a fast subcritical core made of 30% U-235 enriched metallic Uranium in a lead matrix. Several off-line and on-line reactivity measurement techniques will be investigated during the experimental campaign. This report is focused on the simulation by deterministic (ERANOS French code) and Monte Carlo (MCNPX US code) calculations of three reactivity measurement techniques, Slope ({alpha}-fitting), Area-ratio and Source-jerk, applied to a GUINEVERE subcritical configuration (namely SC1). The inferred reactivity, in dollar units, by the Area-ratio method shows an overall agreement between the two deterministic and Monte Carlo computational approaches, whereas the MCNPX Source-jerk results are affected by large uncertainties and allow only partial conclusions about the comparison. Finally, no particular spatial dependence of the results is observed in the case of the GUINEVERE SC1 subcritical configuration. (authors)
Parallel Grand Canonical Monte Carlo (ParaGrandMC) Simulation Code
NASA Technical Reports Server (NTRS)
Yamakov, Vesselin I.
2016-01-01
This report provides an overview of the Parallel Grand Canonical Monte Carlo (ParaGrandMC) simulation code. This is a highly scalable parallel FORTRAN code for simulating the thermodynamic evolution of metal alloy systems at the atomic level, and predicting the thermodynamic state, phase diagram, chemical composition and mechanical properties. The code is designed to simulate multi-component alloy systems, predict solid-state phase transformations such as austenite-martensite transformations, precipitate formation, recrystallization, capillary effects at interfaces, surface absorption, etc., which can aid the design of novel metallic alloys. While the software is mainly tailored for modeling metal alloys, it can also be used for other types of solid-state systems, and to some degree for liquid or gaseous systems, including multiphase systems forming solid-liquid-gas interfaces.
NASA Astrophysics Data System (ADS)
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.
Gas bremsstrahlung studies for medium energy electron storage rings using FLUKA Monte Carlo code
NASA Astrophysics Data System (ADS)
Sahani, Prasanta Kumar; Haridas, G.; Sinha, Anil K.; Hannurkar, P. R.
2016-02-01
Gas bremsstrahlung is generated due to the interaction of the stored electron beam with residual gas molecules of the vacuum chamber in a storage ring. As the opening angle of the bremsstrahlung is very small, the scoring area used in Monte Carlo simulation plays a dominant role in evaluating the absorbed dose. In the present work gas bremsstrahlung angular distribution and absorbed dose for the energies ranging from 1 to 5 GeV electron storage rings are studied using the Monte Carlo code, FLUKA. From the study, an empirical formula for gas bremsstrahlung dose estimation was deduced. The results were compared with the data obtained from reported experimental values. The results obtained from simulations are found to be in very good agreement with the reported experimental data. The results obtained are applied in estimating the gas bremsstrahlung dose for 2.5 GeV synchrotron radiation source, Indus-2 at Raja Ramanna Centre for Advanced Technology, India. The paper discusses the details of the simulation and the results obtained.
Uncertainty analysis in environmental radioactivity measurements using the Monte Carlo code MCNP5
NASA Astrophysics Data System (ADS)
Gallardo, S.; Querol, A.; Ortiz, J.; Ródenas, J.; Verdú, G.; Villanueva, J. F.
2015-11-01
High Purity Germanium (HPGe) detectors are widely used for environmental radioactivity measurements due to their excellent energy resolution. Monte Carlo (MC) codes are a useful tool to complement experimental measurements in calibration procedures at the laboratory. However, the efficiency curve of the detector can vary due to uncertainties associated with measurements. These uncertainties can be classified into some categories: geometrical parameters of the measurement (distance source-detector, volume of the source), properties of the radiation source (radionuclide activity, branching ratio), and detector characteristics (Ge dead layer, active volume, end cap thickness). The Monte Carlo simulation can be also affected by other kind of uncertainties mainly related to cross sections and to the calculation itself. Normally, all these uncertainties are not well known and it required a deep analysis to determine their effect on the detector efficiency. In this work, the Noether-Wilks formula is used to carry out the uncertainty analysis. A Probability Density Function (PDF) is assigned to each variable involved in the sampling process. The size of the sampling is determined from the characteristics of the tolerance intervals by applying the Noether-Wilks formula. Results of the analysis transform the efficiency curve into a region of possible values into the tolerance intervals. Results show a good agreement between experimental measurements and simulations for two different matrices (water and sand).
NASA Astrophysics Data System (ADS)
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
Papadimitroulas, Panagiotis; Loudos, George; Nikiforidis, George C.; Kagadis, George C.
2012-08-15
Purpose: GATE is a Monte Carlo simulation toolkit based on the Geant4 package, widely used for many medical physics applications, including SPECT and PET image simulation and more recently CT image simulation and patient dosimetry. The purpose of the current study was to calculate dose point kernels (DPKs) using GATE, compare them against reference data, and finally produce a complete dataset of the total DPKs for the most commonly used radionuclides in nuclear medicine. Methods: Patient-specific absorbed dose calculations can be carried out using Monte Carlo simulations. The latest version of GATE extends its applications to Radiotherapy and Dosimetry. Comparison of the proposed method for the generation of DPKs was performed for (a) monoenergetic electron sources, with energies ranging from 10 keV to 10 MeV, (b) beta emitting isotopes, e.g., {sup 177}Lu, {sup 90}Y, and {sup 32}P, and (c) gamma emitting isotopes, e.g., {sup 111}In, {sup 131}I, {sup 125}I, and {sup 99m}Tc. Point isotropic sources were simulated at the center of a sphere phantom, and the absorbed dose was stored in concentric spherical shells around the source. Evaluation was performed with already published studies for different Monte Carlo codes namely MCNP, EGS, FLUKA, ETRAN, GEPTS, and PENELOPE. A complete dataset of total DPKs was generated for water (equivalent to soft tissue), bone, and lung. This dataset takes into account all the major components of radiation interactions for the selected isotopes, including the absorbed dose from emitted electrons, photons, and all secondary particles generated from the electromagnetic interactions. Results: GATE comparison provided reliable results in all cases (monoenergetic electrons, beta emitting isotopes, and photon emitting isotopes). The observed differences between GATE and other codes are less than 10% and comparable to the discrepancies observed among other packages. The produced DPKs are in very good agreement with the already published data
Domain Decomposition of a Constructive Solid Geometry Monte Carlo Transport Code
O'Brien, M J; Joy, K I; Procassini, R J; Greenman, G M
2008-12-07
Domain decomposition has been implemented in a Constructive Solid Geometry (CSG) Monte Carlo neutron transport code. Previous methods to parallelize a CSG code relied entirely on particle parallelism; but in our approach we distribute the geometry as well as the particles across processors. This enables calculations whose geometric description is larger than what could fit in memory of a single processor, thus it must be distributed across processors. In addition to enabling very large calculations, we show that domain decomposition can speed up calculations compared to particle parallelism alone. We also show results of a calculation of the proposed Laser Inertial-Confinement Fusion-Fission Energy (LIFE) facility, which has 5.6 million CSG parts.
A portable, parallel, object-oriented Monte Carlo neutron transport code in C++
Lee, S.R.; Cummings, J.C.; Nolen, S.D. |
1997-05-01
We have developed a multi-group Monte Carlo neutron transport code using C++ and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and {alpha}-eigenvalues and is portable to and runs parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities of MC++ are discussed, along with physics and performance results on a variety of hardware, including all Accelerated Strategic Computing Initiative (ASCI) hardware. Current parallel performance indicates the ability to compute {alpha}-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed.
Implementation of the probability table method in a continuous-energy Monte Carlo code system
Sutton, T.M.; Brown, F.B.
1998-10-01
RACER is a particle-transport Monte Carlo code that utilizes a continuous-energy treatment for neutrons and neutron cross section data. Until recently, neutron cross sections in the unresolved resonance range (URR) have been treated in RACER using smooth, dilute-average representations. This paper describes how RACER has been modified to use probability tables to treat cross sections in the URR, and the computer codes that have been developed to compute the tables from the unresolved resonance parameters contained in ENDF/B data files. A companion paper presents results of Monte Carlo calculations that demonstrate the effect of the use of probability tables versus the use of dilute-average cross sections for the URR. The next section provides a brief review of the probability table method as implemented in the RACER system. The production of the probability tables for use by RACER takes place in two steps. The first step is the generation of probability tables from the nuclear parameters contained in the ENDF/B data files. This step, and the code written to perform it, are described in Section 3. The tables produced are at energy points determined by the ENDF/B parameters and/or accuracy considerations. The tables actually used in the RACER calculations are obtained in the second step from those produced in the first. These tables are generated at energy points specific to the RACER calculation. Section 4 describes this step and the code written to implement it, as well as modifications made to RACER to enable it to use the tables. Finally, some results and conclusions are presented in Section 5.
Coded aperture coherent scatter imaging for breast cancer detection: a Monte Carlo evaluation
NASA Astrophysics Data System (ADS)
Lakshmanan, Manu N.; Morris, Robert E.; Greenberg, Joel A.; Samei, Ehsan; Kapadia, Anuj J.
2016-03-01
It is known that conventional x-ray imaging provides a maximum contrast between cancerous and healthy fibroglandular breast tissues of 3% based on their linear x-ray attenuation coefficients at 17.5 keV, whereas coherent scatter signal provides a maximum contrast of 19% based on their differential coherent scatter cross sections. Therefore in order to exploit this potential contrast, we seek to evaluate the performance of a coded- aperture coherent scatter imaging system for breast cancer detection and investigate its accuracy using Monte Carlo simulations. In the simulations we modeled our experimental system, which consists of a raster-scanned pencil beam of x-rays, a bismuth-tin coded aperture mask comprised of a repeating slit pattern with 2-mm periodicity, and a linear-array of 128 detector pixels with 6.5-keV energy resolution. The breast tissue that was scanned comprised a 3-cm sample taken from a patient-based XCAT breast phantom containing a tomosynthesis- based realistic simulated lesion. The differential coherent scatter cross section was reconstructed at each pixel in the image using an iterative reconstruction algorithm. Each pixel in the reconstructed image was then classified as being either air or the type of breast tissue with which its normalized reconstructed differential coherent scatter cross section had the highest correlation coefficient. Comparison of the final tissue classification results with the ground truth image showed that the coded aperture imaging technique has a cancerous pixel detection sensitivity (correct identification of cancerous pixels), specificity (correctly ruling out healthy pixels as not being cancer) and accuracy of 92.4%, 91.9% and 92.0%, respectively. Our Monte Carlo evaluation of our experimental coded aperture coherent scatter imaging system shows that it is able to exploit the greater contrast available from coherently scattered x-rays to increase the accuracy of detecting cancerous regions within the breast.
Extension of the Integrated Tiger Series (ITS) of electron-photon Monte Carlo codes to 100 GeV
Miller, S.G.
1988-08-01
Version 2.1 of the Integrated Tiger Series (ITS) of electron-photon Monte Carlo codes was modified to extend their ability to model interactions up to 100 GeV. Benchmarks against experimental results conducted at 10 and 15 GeV confirm the accuracy of the extended codes. 12 refs., 2 figs., 2 tabs.
SU-E-T-578: MCEBRT, A Monte Carlo Code for External Beam Treatment Plan Verifications
Chibani, O; Ma, C; Eldib, A
2014-06-01
Purpose: Present a new Monte Carlo code (MCEBRT) for patient-specific dose calculations in external beam radiotherapy. The code MLC model is benchmarked and real patient plans are re-calculated using MCEBRT and compared with commercial TPS. Methods: MCEBRT is based on the GEPTS system (Med. Phys. 29 (2002) 835–846). Phase space data generated for Varian linac photon beams (6 – 15 MV) are used as source term. MCEBRT uses a realistic MLC model (tongue and groove, rounded ends). Patient CT and DICOM RT files are used to generate a 3D patient phantom and simulate the treatment configuration (gantry, collimator and couch angles; jaw positions; MLC sequences; MUs). MCEBRT dose distributions and DVHs are compared with those from TPS in absolute way (Gy). Results: Calculations based on the developed MLC model closely matches transmission measurements (pin-point ionization chamber at selected positions and film for lateral dose profile). See Fig.1. Dose calculations for two clinical cases (whole brain irradiation with opposed beams and lung case with eight fields) are carried out and outcomes are compared with the Eclipse AAA algorithm. Good agreement is observed for the brain case (Figs 2-3) except at the surface where MCEBRT dose can be higher by 20%. This is due to better modeling of electron contamination by MCEBRT. For the lung case an overall good agreement (91% gamma index passing rate with 3%/3mm DTA criterion) is observed (Fig.4) but dose in lung can be over-estimated by up to 10% by AAA (Fig.5). CTV and PTV DVHs from TPS and MCEBRT are nevertheless close (Fig.6). Conclusion: A new Monte Carlo code is developed for plan verification. Contrary to phantombased QA measurements, MCEBRT simulate the exact patient geometry and tissue composition. MCEBRT can be used as extra verification layer for plans where surface dose and tissue heterogeneity are an issue.
Development of a Space Radiation Monte-Carlo Computer Simulation Based on the FLUKE and Root Codes
NASA Technical Reports Server (NTRS)
Pinsky, L. S.; Wilson, T. L.; Ferrari, A.; Sala, Paola; Carminati, F.; Brun, R.
2001-01-01
The radiation environment in space is a complex problem to model. Trying to extrapolate the projections of that environment into all areas of the internal spacecraft geometry is even more daunting. With the support of our CERN colleagues, our research group in Houston is embarking on a project to develop a radiation transport tool that is tailored to the problem of taking the external radiation flux incident on any particular spacecraft and simulating the evolution of that flux through a geometrically accurate model of the spacecraft material. The output will be a prediction of the detailed nature of the resulting internal radiation environment within the spacecraft as well as its secondary albedo. Beyond doing the physics transport of the incident flux, the software tool we are developing will provide a self-contained stand-alone object-oriented analysis and visualization infrastructure. It will also include a graphical user interface and a set of input tools to facilitate the simulation of space missions in terms of nominal radiation models and mission trajectory profiles. The goal of this project is to produce a code that is considerably more accurate and user-friendly than existing Monte-Carlo-based tools for the evaluation of the space radiation environment. Furthermore, the code will be an essential complement to the currently existing analytic codes in the BRYNTRN/HZETRN family for the evaluation of radiation shielding. The code will be directly applicable to the simulation of environments in low earth orbit, on the lunar surface, on planetary surfaces (including the Earth) and in the interplanetary medium such as on a transit to Mars (and even in the interstellar medium). The software will include modules whose underlying physics base can continue to be enhanced and updated for physics content, as future data become available beyond the timeframe of the initial development now foreseen. This future maintenance will be available from the authors of FLUKA as
TRIPOLI-4®, CEA, EDF and AREVA Reference Monte Carlo Code
NASA Astrophysics Data System (ADS)
2014-06-01
This paper presents an overview of TRIPOLI-4®, the fourth generation of the 3D continuous-energy Monte Carlo code developed by the Service d'Etudes des Réacteurs et de Mathématiques Appliquées (SERMA) at CEA Saclay. The paper surveys the generic features: programming language, parallel operation, tracked particles, nuclear data, geometry, simulation modes, standard variance reduction techniques, sources, tracking and collision algorithms, tallies, sensitivity studies. Moreover, specific and recent features are also detailed: Doppler broadening of the elastic scattering kernel, neutron and photon material irradiation, advanced variance reduction techniques, Green's functions, cycle correlation correction, nuclear data management and depletion capabilities. The productivity tools (T4G, SALOME TRIPOLI, T4RootTools), the Verification & Validation process and the distribution and licensing policy are finally presented.
Space applications of the MITS electron-photon Monte Carlo transport code system
Kensek, R.P.; Lorence, L.J.; Halbleib, J.A.; Morel, J.E.
1996-07-01
The MITS multigroup/continuous-energy electron-photon Monte Carlo transport code system has matured to the point that it is capable of addressing more realistic three-dimensional adjoint applications. It is first employed to efficiently predict point doses as a function of source energy for simple three-dimensional experimental geometries exposed to simulated uniform isotropic planar sources of monoenergetic electrons up to 4.0 MeV. Results are in very good agreement with experimental data. It is then used to efficiently simulate dose to a detector in a subsystem of a GPS satellite due to its natural electron environment, employing a relatively complex model of the satellite. The capability for survivability analysis of space systems is demonstrated, and results are obtained with and without variance reduction.
Zhao, L.; Cluggish, B.; Kim, J. S.; Pardo, R.; Vondrasek, R.
2010-02-15
A Monte Carlo charge breeding code (MCBC) is being developed by FAR-TECH, Inc. to model the capture and charge breeding of 1+ ion beam in an electron cyclotron resonance ion source (ECRIS) device. The ECRIS plasma is simulated using the generalized ECRIS model which has two choices of boundary settings, free boundary condition and Bohm condition. The charge state distribution of the extracted beam ions is calculated by solving the steady state ion continuity equations where the profiles of the captured ions are used as source terms. MCBC simulations of the charge breeding of Rb+ showed good agreement with recent charge breeding experiments at Argonne National Laboratory (ANL). MCBC correctly predicted the peak of highly charged ion state outputs under free boundary condition and similar charge state distribution width but a lower peak charge state under the Bohm condition. The comparisons between the simulation results and ANL experimental measurements are presented and discussed.
Full modelling of the MOSAIC animal PET system based on the GATE Monte Carlo simulation code
NASA Astrophysics Data System (ADS)
Merheb, C.; Petegnief, Y.; Talbot, J. N.
2007-02-01
Positron emission tomography (PET) systems dedicated to animal imaging are now widely used for biological studies. The scanner performance strongly depends on the design and the characteristics of the system. Many parameters must be optimized like the dimensions and type of crystals, geometry and field-of-view (FOV), sampling, electronics, lightguide, shielding, etc. Monte Carlo modelling is a powerful tool to study the effect of each of these parameters on the basis of realistic simulated data. Performance assessment in terms of spatial resolution, count rates, scatter fraction and sensitivity is an important prerequisite before the model can be used instead of real data for a reliable description of the system response function or for optimization of reconstruction algorithms. The aim of this study is to model the performance of the Philips Mosaic™ animal PET system using a comprehensive PET simulation code in order to understand and describe the origin of important factors that influence image quality. We use GATE, a Monte Carlo simulation toolkit for a realistic description of the ring PET model, the detectors, shielding, cap, electronic processing and dead times. We incorporate new features to adjust signal processing to the Anger logic underlying the Mosaic™ system. Special attention was paid to dead time and energy spectra descriptions. Sorting of simulated events in a list mode format similar to the system outputs was developed to compare experimental and simulated sensitivity and scatter fractions for different energy thresholds using various models of phantoms describing rat and mouse geometries. Count rates were compared for both cylindrical homogeneous phantoms. Simulated spatial resolution was fitted to experimental data for 18F point sources at different locations within the FOV with an analytical blurring function for electronic processing effects. Simulated and measured sensitivities differed by less than 3%, while scatter fractions agreed
Giuseppe Palmiotti
2015-05-01
In this work, the implementation of a collision history-based approach to sensitivity/perturbation calculations in the Monte Carlo code SERPENT is discussed. The proposed methods allow the calculation of the eects of nuclear data perturbation on several response functions: the eective multiplication factor, reaction rate ratios and bilinear ratios (e.g., eective kinetics parameters). SERPENT results are compared to ERANOS and TSUNAMI Generalized Perturbation Theory calculations for two fast metallic systems and for a PWR pin-cell benchmark. New methods for the calculation of sensitivities to angular scattering distributions are also presented, which adopts fully continuous (in energy and angle) Monte Carlo estimators.
NASA Astrophysics Data System (ADS)
Ilgüsatiroglu, Emre; Illarionov, Alexey Yu.; Ciappa, Mauro; Pfäffli, Paul; Bomholt, Lars
2014-04-01
A new Monte Carlo code is presented that includes among others definition of arbitrary geometries with sub-nanometer resolution, high performance parallel computing capabilities, trapped charge, electric field calculation, electron tracking in electrostatic field, and calculation of 3D dose distributions. These functionalities are efficiently implemented thanks to the coupling of the Monte Carlo simulator with a TCAD environment. Applications shown are the synthesis of SEM linescans and images that focus on the evaluation of the impact of proximity effects and self charging on the quantitative extraction of critical dimensions in dense photoresist structures.
Juste, B; Miro, R; Gallardo, S; Santos, A; Verdu, G
2006-01-01
The present work has simulated the photon and electron transport in a Theratron 780 (MDS Nordion) (60)Co radiotherapy unit, using the Monte Carlo transport code, MCNP (Monte Carlo N-Particle), version 5. In order to become computationally more efficient in view of taking part in the practical field of radiotherapy treatment planning, this work is focused mainly on the analysis of dose results and on the required computing time of different tallies applied in the model to speed up calculations. PMID:17946330
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.
A Comparison Between GATE and MCNPX Monte Carlo Codes in Simulation of Medical Linear Accelerator
Sadoughi, Hamid-Reza; Nasseri, Shahrokh; Momennezhad, Mahdi; Sadeghi, Hamid-Reza; Bahreyni-Toosi, Mohammad-Hossein
2014-01-01
Radiotherapy dose calculations can be evaluated by Monte Carlo (MC) simulations with acceptable accuracy for dose prediction in complicated treatment plans. In this work, Standard, Livermore and Penelope electromagnetic (EM) physics packages of GEANT4 application for tomographic emission (GATE) 6.1 were compared versus Monte Carlo N-Particle eXtended (MCNPX) 2.6 in simulation of 6 MV photon Linac. To do this, similar geometry was used for the two codes. The reference values of percentage depth dose (PDD) and beam profiles were obtained using a 6 MV Elekta Compact linear accelerator, Scanditronix water phantom and diode detectors. No significant deviations were found in PDD, dose profile, energy spectrum, radial mean energy and photon radial distribution, which were calculated by Standard and Livermore EM models and MCNPX, respectively. Nevertheless, the Penelope model showed an extreme difference. Statistical uncertainty in all the simulations was <1%, namely 0.51%, 0.27%, 0.27% and 0.29% for PDDs of 10 cm2× 10 cm2 filed size, for MCNPX, Standard, Livermore and Penelope models, respectively. Differences between spectra in various regions, in radial mean energy and in photon radial distribution were due to different cross section and stopping power data and not the same simulation of physics processes of MCNPX and three EM models. For example, in the Standard model, the photoelectron direction was sampled from the Gavrila-Sauter distribution, but the photoelectron moved in the same direction of the incident photons in the photoelectric process of Livermore and Penelope models. Using the same primary electron beam, the Standard and Livermore EM models of GATE and MCNPX showed similar output, but re-tuning of primary electron beam is needed for the Penelope model. PMID:24696804
Verification of SMART Neutronics Design Methodology by the MCNAP Monte Carlo Code
Jong Sung Chung; Kyung Jin Shim; Chang Hyo Kim; Chungchan Lee; Sung Quun Zee
2000-11-12
SMART is a small advanced integral pressurized water reactor (PWR) of 330 MW(thermal) designed for both electricity generation and seawater desalinization. The CASMO-3/MASTER nuclear analysis system, a design-basis of Korean PWR plants, has been employed for the SMART core nuclear design and analysis because the fuel assembly (FA) characteristics and reactor operating conditions in temperature and pressure are similar to those of PWR plants. However, the SMART FAs are highly poisoned with more than 20 Al{sub 2}O{sub 3}-B{sub 4}C plus additional Gd{sub 2}O{sub 3}/UO{sub 2} BPRs each FA. The reactor is operated with control rods inserted. Therefore, the flux and power distribution may become more distorted than those of commercial PWR plants. In addition, SMART should produce power from room temperature to hot-power operating condition because it employs nuclear heating from room temperature. This demands reliable predictions of core criticality, shutdown margin, control rod worth, power distributions, and reactivity coefficients at both room temperature and hot operating condition, yet no such data are available to verify the CASMO-3/MASTER (hereafter MASTER) code system. In the absence of experimental verification data for the SMART neutronics design, the Monte Carlo depletion analysis program MCNAP is adopted as near-term alternatives for qualifying MASTER neutronics design calculations. The MCNAP is a personal computer-based continuous energy Monte Carlo neutronics analysis program written in C++ language. We established its qualification by presenting its prediction accuracy on measurements of Venus critical facilities and core neutronics analysis of a PWR plant in operation, and depletion characteristics of integral burnable absorber FAs of the current PWR. Here, we present a comparison of MASTER and MCNAP neutronics design calculations for SMART and establish the qualification of the MASTER system.
Energy Science and Technology Software Center (ESTSC)
2013-06-24
Version 07 TART2012 is a coupled neutron-photon Monte Carlo transport code designed to use three-dimensional (3-D) combinatorial geometry. Neutron and/or photon sources as well as neutron induced photon production can be tracked. It is a complete system to assist you with input preparation, running Monte Carlo calculations, and analysis of output results. TART2012 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared tomore » other similar codes. Use of the entire system can save you a great deal of time and energy. TART2012 extends the general utility of the code to even more areas of application than available in previous releases by concentrating on improving the physics, particularly with regard to improved treatment of neutron fission, resonance self-shielding, molecular binding, and extending input options used by the code. Several utilities are included for creating input files and displaying TART results and data. TART2012 uses the latest ENDF/B-VI, Release 8, data. New for TART2012 is the use of continuous energy neutron cross sections, in addition to its traditional multigroup cross sections. For neutron interaction, the data are derived using ENDF-ENDL2005 and include both continuous energy cross sections and 700 group neutron data derived using a combination of ENDF/B-VI, Release 8, and ENDL data. The 700 group structure extends from 10-5 eV up to 1 GeV. Presently nuclear data are only available up to 20 MeV, so that only 616 of the groups are currently used. For photon interaction, 701 point photon data were derived using the Livermore EPDL97 file. The new 701 point structure extends from 100 eV up to 1 GeV, and is currently used over this entire energy range. TART2012 completely supersedes all older versions of TART, and it is strongly recommended that one use only the most recent version of TART2012 and its data files. Check authors homepage for related information: http
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.
NASA Astrophysics Data System (ADS)
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.
Computation of a Canadian SCWR unit cell with deterministic and Monte Carlo codes
Harrisson, G.; Marleau, G.
2012-07-01
The Canadian SCWR has the potential to achieve the goals that the generation IV nuclear reactors must meet. As part of the optimization process for this design concept, lattice cell calculations are routinely performed using deterministic codes. In this study, the first step (self-shielding treatment) of the computation scheme developed with the deterministic code DRAGON for the Canadian SCWR has been validated. Some options available in the module responsible for the resonance self-shielding calculation in DRAGON 3.06 and different microscopic cross section libraries based on the ENDF/B-VII.0 evaluated nuclear data file have been tested and compared to a reference calculation performed with the Monte Carlo code SERPENT under the same conditions. Compared to SERPENT, DRAGON underestimates the infinite multiplication factor in all cases. In general, the original Stammler model with the Livolant-Jeanpierre approximations are the most appropriate self-shielding options to use in this case of study. In addition, the 89 groups WIMS-AECL library for slight enriched uranium and the 172 groups WLUP library for a mixture of plutonium and thorium give the most consistent results with those of SERPENT. (authors)
Mesh-based Monte Carlo code for fluorescence modeling in complex tissues with irregular boundaries
NASA Astrophysics Data System (ADS)
Wilson, Robert H.; Chen, Leng-Chun; Lloyd, William; Kuo, Shiuhyang; Marcelo, Cynthia; Feinberg, Stephen E.; Mycek, Mary-Ann
2011-07-01
There is a growing need for the development of computational models that can account for complex tissue morphology in simulations of photon propagation. We describe the development and validation of a user-friendly, MATLAB-based Monte Carlo code that uses analytically-defined surface meshes to model heterogeneous tissue geometry. The code can use information from non-linear optical microscopy images to discriminate the fluorescence photons (from endogenous or exogenous fluorophores) detected from different layers of complex turbid media. We present a specific application of modeling a layered human tissue-engineered construct (Ex Vivo Produced Oral Mucosa Equivalent, EVPOME) designed for use in repair of oral tissue following surgery. Second-harmonic generation microscopic imaging of an EVPOME construct (oral keratinocytes atop a scaffold coated with human type IV collagen) was employed to determine an approximate analytical expression for the complex shape of the interface between the two layers. This expression can then be inserted into the code to correct the simulated fluorescence for the effect of the irregular tissue geometry.
SU-E-T-323: The FLUKA Monte Carlo Code in Ion Beam Therapy
Rinaldi, I
2014-06-01
Purpose: Monte Carlo (MC) codes are increasingly used in the ion beam therapy community due to their detailed description of radiation transport and interaction with matter. The suitability of a MC code demands accurate and reliable physical models for the transport and the interaction of all components of the mixed radiation field. This contribution will address an overview of the recent developments in the FLUKA code oriented to its application in ion beam therapy. Methods: FLUKA is a general purpose MC code which allows the calculations of particle transport and interactions with matter, covering an extended range of applications. The user can manage the code through a graphic interface (FLAIR) developed using the Python programming language. Results: This contribution will present recent refinements in the description of the ionization processes and comparisons between FLUKA results and experimental data of ion beam therapy facilities. Moreover, several validations of the largely improved FLUKA nuclear models for imaging application to treatment monitoring will be shown. The complex calculation of prompt gamma ray emission compares favorably with experimental data and can be considered adequate for the intended applications. New features in the modeling of proton induced nuclear interactions also provide reliable cross section predictions for the production of radionuclides. Of great interest for the community are the developments introduced in FLAIR. The most recent efforts concern the capability of importing computed-tomography images in order to build automatically patient geometries and the implementation of different types of existing positron-emission-tomography scanner devices for imaging applications. Conclusion: The FLUA code has been already chosen as reference MC code in many ion beam therapy centers, and is being continuously improved in order to match the needs of ion beam therapy applications. Parts of this work have been supported by the European
Assessment of MIRD data for internal dosimetry using the GATE Monte Carlo code.
Parach, Ali Asghar; Rajabi, Hossein; Askari, Mohammad Ali
2011-08-01
GATE/GEANT is a Monte Carlo code dedicated to nuclear medicine that allows calculation of the dose to organs of voxel phantoms. On the other hand, MIRD is a well-developed system for estimation of the dose to human organs. In this study, results obtained from GATE/GEANT using Snyder phantom are compared to published MIRD data. For this, the mathematical Snyder phantom was discretized and converted to a digital phantom of 100 × 200 × 360 voxels. The activity was considered uniformly distributed within kidneys, liver, lungs, pancreas, spleen, and adrenals. The GATE/GEANT Monte Carlo code was used to calculate the dose to the organs of the phantom from mono-energetic photons of 10, 15, 20, 30, 50, 100, 200, 500, and 1000 keV. The dose was converted into specific absorbed fraction (SAF) and the results were compared to the corresponding published MIRD data. On average, there was a good correlation (r (2)>0.99) between the two series of data. However, the GATE/GEANT data were on average -0.16 ± 6.22% lower than the corresponding MIRD data for self-absorption. Self-absorption in the lungs was considerably higher in the MIRD compared to the GATE/GEANT data, for photon energies of 10-20 keV. As for cross-irradiation to other organs, the GATE/GEANT data were on average +1.5 ± 8.1% higher than the MIRD data, for photon energies of 50-1000 keV. For photon energies of 10-30 keV, the relative difference was +7.5 ± 67%. It turned out that the agreement between the GATE/GEANT and the MIRD data depended upon absolute SAF values and photon energy. For 10-30 keV photons, where the absolute SAF values were small, the uncertainty was high and the effect of cross-section prominent, and there was no agreement between the GATE/GEANT results and the MIRD data. However, for photons of 50-1,000 keV, the bias was negligible and the agreement was acceptable. PMID:21573984
NASA Astrophysics Data System (ADS)
Brogan, John
Understanding the dosimetry for high-energy, heavy ions (HZE), especially within living systems, is complex and requires the use of both experimental and computational methods. Tissue-equivalent proportional counters (TEPCs) have been used experimentally to measure energy deposition in volumes similar in dimension to a mammalian cell. As these experiments begin to include a wider range of ions and energies, considerations to cost, time, and radiation protection are necessary and may limit the extent of these studies. Multiple Monte Carlo computational codes have been created to remediate this problem and serve as a mode of verification for pervious experimental methods. One such code, Relativistic-Ion Tracks (RITRACKS), is currently being developed at the NASA Johnson Space center. RITRACKS was designed to describe patterns of ionizations responsible for DNA damage on the molecular scale (nanometers). This study extends RITRACKS version 3.07 into the microdosimetric scale (microns), and compares computational results to previous experimental TEPC data. Energy deposition measurements for 1000 MeV nucleon-1 Fe ions in a 1 micron spherical target were compared. Different settings within RITRACKS were tested to verify their effects on dose to a target and the resulting energy deposition frequency distribution. The results were then compared to the TEPC data.
ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®
NASA Astrophysics Data System (ADS)
Damian, F.; Brun, E.
2014-06-01
ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14 MW. ORPHEE core is 90 cm height and has a cross section of 27x27 cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4® [1,2,3]. A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 [4]. The analysis showed the reliability of TRIPOLI-4® to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.
Integrated TIGER Series of Coupled Electron/Photon Monte Carlo Transport Codes System.
Energy Science and Technology Software Center (ESTSC)
2012-11-30
Version: 00 Distribution is restricted to US Government Agencies and Their Contractors Only. The Integrated Tiger Series (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. The goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects onemore » 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 95. 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.« less
Integrated TIGER Series of Coupled Electron/Photon Monte Carlo Transport Codes System.
VALDEZ, GREG D.
2012-11-30
Version: 00 Distribution is restricted to US Government Agencies and Their Contractors Only. The Integrated Tiger Series (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. The 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 95. 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.
Procassini, R.J.
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.
MOCRA: a Monte Carlo code for the simulation of radiative transfer in the atmosphere.
Premuda, Margherita; Palazzi, Elisa; Ravegnani, Fabrizio; Bortoli, Daniele; Masieri, Samuele; Giovanelli, Giorgio
2012-03-26
This paper describes the radiative transfer model (RTM) MOCRA (MOnte Carlo Radiance Analysis), developed in the frame of DOAS (Differential Optical Absorption Spectroscopy) to correctly interpret remote sensing measurements of trace gas amounts in the atmosphere through the calculation of the Air Mass Factor. Besides the DOAS-related quantities, the MOCRA code yields: 1- the atmospheric transmittance in the vertical and sun directions, 2- the direct and global irradiance, 3- the single- and multiple- scattered radiance for a detector with assigned position, line of sight and field of view. Sample calculations of the main radiometric quantities calculated with MOCRA are presented and compared with the output of another RTM (MODTRAN4). A further comparison is presented between the NO2 slant column densities (SCDs) measured with DOAS at Evora (Portugal) and the ones simulated with MOCRA. Both comparisons (MOCRA-MODTRAN4 and MOCRA-observations) gave more than satisfactory results, and overall make MOCRA a versatile tool for atmospheric radiative transfer simulations and interpretation of remote sensing measurements. PMID:22453470
Code System for Monte Carlo Simulation of Electron and Photon Transport.
2015-07-01
Version 01 PENELOPE performs Monte Carlo simulation of coupled electron-photon transport in arbitrary materials and complex quadric geometries. A mixed procedure is used for the simulation of electron and positron interactions (elastic scattering, inelastic scattering and bremsstrahlung emission), in which hard events (i.e. those with deflection angle and/or energy loss larger than pre-selected cutoffs) are simulated in a detailed way, while soft interactions are calculated from multiple scattering approaches. Photon interactions (Rayleigh scattering, Compton scattering, photoelectric effect and electron-positron pair production) and positron annihilation are simulated in a detailed way. PENELOPE reads the required physical information about each material (which includes tables of physical properties, interaction cross sections, relaxation data, etc.) from the input material data file. The material data file is created by means of the auxiliary program MATERIAL, which extracts atomic interaction data from the database of ASCII files. PENELOPE mailing list archives and additional information about the code can be found at http://www.nea.fr/lists/penelope.html. See Abstract for additional features.
Code System for Monte Carlo Simulation of Electron and Photon Transport.
Energy Science and Technology Software Center (ESTSC)
2015-07-01
Version 01 PENELOPE performs Monte Carlo simulation of coupled electron-photon transport in arbitrary materials and complex quadric geometries. A mixed procedure is used for the simulation of electron and positron interactions (elastic scattering, inelastic scattering and bremsstrahlung emission), in which hard events (i.e. those with deflection angle and/or energy loss larger than pre-selected cutoffs) are simulated in a detailed way, while soft interactions are calculated from multiple scattering approaches. Photon interactions (Rayleigh scattering, Compton scattering,more » photoelectric effect and electron-positron pair production) and positron annihilation are simulated in a detailed way. PENELOPE reads the required physical information about each material (which includes tables of physical properties, interaction cross sections, relaxation data, etc.) from the input material data file. The material data file is created by means of the auxiliary program MATERIAL, which extracts atomic interaction data from the database of ASCII files. PENELOPE mailing list archives and additional information about the code can be found at http://www.nea.fr/lists/penelope.html. See Abstract for additional features.« less
A PARALLEL MONTE CARLO CODE FOR SIMULATING COLLISIONAL N-BODY SYSTEMS
Pattabiraman, Bharath; Umbreit, Stefan; Liao, Wei-keng; Choudhary, Alok; Kalogera, Vassiliki; Memik, Gokhan; Rasio, Frederic A.
2013-02-15
We present a new parallel code for computing the dynamical evolution of collisional N-body systems with up to N {approx} 10{sup 7} particles. Our code is based on the Henon Monte Carlo method for solving the Fokker-Planck equation, and makes assumptions of spherical symmetry and dynamical equilibrium. The principal algorithmic developments involve optimizing data structures and the introduction of a parallel random number generation scheme as well as a parallel sorting algorithm required to find nearest neighbors for interactions and to compute the gravitational potential. The new algorithms we introduce along with our choice of decomposition scheme minimize communication costs and ensure optimal distribution of data and workload among the processing units. Our implementation uses the Message Passing Interface library for communication, which makes it portable to many different supercomputing architectures. We validate the code by calculating the evolution of clusters with initial Plummer distribution functions up to core collapse with the number of stars, N, spanning three orders of magnitude from 10{sup 5} to 10{sup 7}. We find that our results are in good agreement with self-similar core-collapse solutions, and the core-collapse times generally agree with expectations from the literature. Also, we observe good total energy conservation, within {approx}< 0.04% throughout all simulations. We analyze the performance of the code, and demonstrate near-linear scaling of the runtime with the number of processors up to 64 processors for N = 10{sup 5}, 128 for N = 10{sup 6} and 256 for N = 10{sup 7}. The runtime reaches saturation with the addition of processors beyond these limits, which is a characteristic of the parallel sorting algorithm. The resulting maximum speedups we achieve are approximately 60 Multiplication-Sign , 100 Multiplication-Sign , and 220 Multiplication-Sign , respectively.
NASA Astrophysics Data System (ADS)
SU, J.; Sagdeev, R.; Usikov, D.; Chin, G.; Boyer, L.; Livengood, T. A.; McClanahan, T. P.; Murray, J.; Starr, R. D.
2013-12-01
Introduction: The leakage flux of lunar neutrons produced by precipitation of galactic cosmic ray (GCR) particles in the upper layer of the lunar regolith and measured by orbital instruments such as the Lunar Exploration Neutron Detector (LEND) is investigated by Monte Carlo simulation. Previous Monte Carlo (MC) simulations have been used to investigate neutron production and leakage from the lunar surface to assess the elemental composition of lunar soil [1-6] and its effect on the leakage neutron flux. We investigate effects on the emergent flux that depend on the physical distribution of hydrogen within the regolith. We use the software package GEANT4 [7] to calculate neutron production from spallation by GCR particles [8,9] in the lunar soil. Multiple layers of differing hydrogen/water at different depths in the lunar regolith model are introduced to examine enhancement or suppression of leakage neutron flux. We find that the majority of leakage thermal and epithermal neutrons are produced in 25 cm to 75 cm deep from the lunar surface. Neutrons produced in the shallow top layer retain more of their original energy due to fewer scattering interactions and escape from the lunar surface mostly as fast neutrons. This provides a diagnostic tool in interpreting leakage neutron flux enhancement or suppression due to hydrogen concentration distribution in lunar regolith. We also find that the emitting angular distribution of thermal and epithermal leakage neutrons can be described by cos3/2(theta) where the fast neutrons emitting angular distribution is cos(theta). The energy sensitivity and angular response of the LEND detectors SETN and CSETN are investigated using the leakage neutron spectrum from GEANT4 simulations. A simplified LRO model is used to benchmark MCNPX[10] and GEANT4 on CSETN absolute count rate corresponding to neutron flux from bombardment of 120MV solar potential GCR particles on FAN lunar soil. We are able to interpret the count rates of SETN and
Marcus, Ryan C.
2012-07-25
MCMini is a proof of concept that demonstrates the possibility for Monte Carlo neutron transport using OpenCL with a focus on performance. This implementation, written in C, shows that tracing particles and calculating reactions on a 3D mesh can be done in a highly scalable fashion. These results demonstrate a potential path forward for MCNP or other Monte Carlo codes.
NASA Technical Reports Server (NTRS)
Shinn, Judy L.; Wilson, John W.; Lone, M. A.; Wong, P. Y.; Costen, Robert C.
1994-01-01
A baryon transport code (BRYNTRN) has previously been verified using available Monte Carlo results for a solar-flare spectrum as the reference. Excellent results were obtained, but the comparisons were limited to the available data on dose and dose equivalent for moderate penetration studies that involve minor contributions from secondary neutrons. To further verify the code, the secondary energy spectra of protons and neutrons are calculated using BRYNTRN and LAHET (Los Alamos High-Energy Transport code, which is a Monte Carlo code). These calculations are compared for three locations within a water slab exposed to the February 1956 solar-proton spectrum. Reasonable agreement was obtained when various considerations related to the calculational techniques and their limitations were taken into account. Although the Monte Carlo results are preliminary, it appears that the neutron albedo, which is not currently treated in BRYNTRN, might be a cause for the large discrepancy seen at small penetration depths. It also appears that the nonelastic neutron production cross sections in BRYNTRN may underestimate the number of neutrons produced in proton collisions with energies below 200 MeV. The notion that the poor energy resolution in BRYNTRN may cause a large truncation error in neutron elastic scattering requires further study.
Parallelizing Monte Carlo with PMC
Rathkopf, J.A.; Jones, T.R.; Nessett, D.M.; Stanberry, L.C.
1994-11-01
PMC (Parallel Monte Carlo) is a system of generic interface routines that allows easy porting of Monte Carlo packages of large-scale physics simulation codes to Massively Parallel Processor (MPP) computers. By loading various versions of PMC, simulation code developers can configure their codes to run in several modes: serial, Monte Carlo runs on the same processor as the rest of the code; parallel, Monte Carlo runs in parallel across many processors of the MPP with the rest of the code running on other MPP processor(s); distributed, Monte Carlo runs in parallel across many processors of the MPP with the rest of the code running on a different machine. This multi-mode approach allows maintenance of a single simulation code source regardless of the target machine. PMC handles passing of messages between nodes on the MPP, passing of messages between a different machine and the MPP, distributing work between nodes, and providing independent, reproducible sequences of random numbers. Several production codes have been parallelized under the PMC system. Excellent parallel efficiency in both the distributed and parallel modes results if sufficient workload is available per processor. Experiences with a Monte Carlo photonics demonstration code and a Monte Carlo neutronics package are described.
Kramer, R; Vieira, J W; Lima, F R A; Fuelle, D
2002-07-01
Organ or tissue equivalent dose, the most important quantity in radiation protection, cannot be measured directly. Therefore it became common practice to calculate the quantity of interest with Monte Carlo methods applied to so-called human phantoms, which are virtual representations of the human body. The Monte Carlo computer code determines conversion coefficients, which are ratios between organ or tissue equivalent dose and measurable quantities. Conversion coefficients have been published by the ICRP (Report No. 74) for various types of radiation, energies and fields, which have been calculated, among others, with the mathematical phantoms ADAM and EVA. Since then progress of image processing, and of clock speed and memory capacity of computers made it possible to create so-called voxel phantoms, which are a far more realistic representation of the human body. Voxel (Volume pixel) phantoms are built from segmented CT and/or MRI images of real persons. A complete set of such images can be joined to a 3-dimensional representation of the human body, which can be linked to a Monte Carlo code allowing for particle transport calculations. A modified version of the VOX_TISS8 human voxel phantom (Yale University) has been connected to the EGS4 Monte Carlo code. The paper explains the modifications, which have been made, the method of coupling the voxel phantom with the code, and presents results as conversion coefficients between organ equivalent dose and kerma in air for external photon radiation. A comparison of the results with published data shows good agreement. PMID:12146699
Valentine, T.E.; Rugama, Y. Munoz-Cobos, J.; Perez, R.
2000-10-23
The design of reactivity monitoring systems for accelerator-driven systems must be investigated to ensure that such systems remain subcritical during operation. The Monte Carlo codes LAHET and MCNP-DSP were combined together to facilitate the design of reactivity monitoring systems. The coupling of LAHET and MCNP-DSP provides a tool that can be used to simulate a variety of subcritical measurements such as the pulsed neutron, Rossi-{alpha}, or noise analysis measurements.
NASA Astrophysics Data System (ADS)
Dytman, Steven
2011-10-01
Every neutrino experiment requires a Monte Carlo event generator for various purposes. Historically, each series of experiments developed their own code which tuned to their needs. Modern experiments would benefit from a universal code (e.g. PYTHIA) which would allow more direct comparison between experiments. GENIE attempts to be that code. This paper compares most commonly used codes and provides some details of GENIE.
Tyagi, Neelam; Bose, Abhijit; Chetty, Indrin J
2004-09-01
We have parallelized the Dose Planning Method (DPM), a Monte Carlo code optimized for radiotherapy class problems, on distributed-memory processor architectures using the Message Passing Interface (MPI). Parallelization has been investigated on a variety of parallel computing architectures at the University of Michigan-Center for Advanced Computing, with respect to efficiency and speedup as a function of the number of processors. We have integrated the parallel pseudo random number generator from the Scalable Parallel Pseudo-Random Number Generator (SPRNG) library to run with the parallel DPM. The Intel cluster consisting of 800 MHz Intel Pentium III processor shows an almost linear speedup up to 32 processors for simulating 1 x 10(8) or more particles. The speedup results are nearly linear on an Athlon cluster (up to 24 processors based on availability) which consists of 1.8 GHz+ Advanced Micro Devices (AMD) Athlon processors on increasing the problem size up to 8 x 10(8) histories. For a smaller number of histories (1 x 10(8)) the reduction of efficiency with the Athlon cluster (down to 83.9% with 24 processors) occurs because the processing time required to simulate 1 x 10(8) histories is less than the time associated with interprocessor communication. A similar trend was seen with the Opteron Cluster (consisting of 1400 MHz, 64-bit AMD Opteron processors) on increasing the problem size. Because of the 64-bit architecture Opteron processors are capable of storing and processing instructions at a faster rate and hence are faster as compared to the 32-bit Athlon processors. We have validated our implementation with an in-phantom dose calculation study using a parallel pencil monoenergetic electron beam of 20 MeV energy. The phantom consists of layers of water, lung, bone, aluminum, and titanium. The agreement in the central axis depth dose curves and profiles at different depths shows that the serial and parallel codes are equivalent in accuracy. PMID:15487756
Icarus: A 2-D Direct Simulation Monte Carlo (DSMC) Code for Multi-Processor Computers
BARTEL, TIMOTHY J.; PLIMPTON, STEVEN J.; GALLIS, MICHAIL A.
2001-10-01
Icarus is a 2D Direct Simulation Monte Carlo (DSMC) code which has been optimized for the parallel computing environment. The code is based on the DSMC method of Bird[11.1] and models from free-molecular to continuum flowfields in either cartesian (x, y) or axisymmetric (z, r) coordinates. Computational particles, representing a given number of molecules or atoms, are tracked as they have collisions with other particles or surfaces. Multiple species, internal energy modes (rotation and vibration), chemistry, and ion transport are modeled. A new trace species methodology for collisions and chemistry is used to obtain statistics for small species concentrations. Gas phase chemistry is modeled using steric factors derived from Arrhenius reaction rates or in a manner similar to continuum modeling. Surface chemistry is modeled with surface reaction probabilities; an optional site density, energy dependent, coverage model is included. Electrons are modeled by either a local charge neutrality assumption or as discrete simulational particles. Ion chemistry is modeled with electron impact chemistry rates and charge exchange reactions. Coulomb collision cross-sections are used instead of Variable Hard Sphere values for ion-ion interactions. The electro-static fields can either be: externally input, a Langmuir-Tonks model or from a Green's Function (Boundary Element) based Poison Solver. Icarus has been used for subsonic to hypersonic, chemically reacting, and plasma flows. The Icarus software package includes the grid generation, parallel processor decomposition, post-processing, and restart software. The commercial graphics package, Tecplot, is used for graphics display. All of the software packages are written in standard Fortran.
A Two-Dimensional Monte Carlo Code System for Linear Neutron Transport Calculations.
Energy Science and Technology Software Center (ESTSC)
1980-04-24
Version 00 KIM (k-infinite-Monte Carlo) solves the steady-state linear neutron transport equation for a fixed source problem or, by successive fixed-source runs, for the eigenvalue problem, in a two-dimensional infinite thermal reactor lattice using the Monte Carlo method. In addition to the combinatorial description of domains, the program allows complex configurations to be represented by a discrete set of points whereby the calculation speed is greatly improved. Configurations are described as the result of overlaysmore » of elementary figures over a basic domain.« less
Somasundaram, E.; Palmer, T. S.
2013-07-01
In this paper, the work that has been done to implement variance reduction techniques in a three dimensional, multi group Monte Carlo code - Tortilla, that works within the frame work of the commercial deterministic code - Attila, is presented. This project is aimed to develop an integrated Hybrid code that seamlessly takes advantage of the deterministic and Monte Carlo methods for deep shielding radiation detection problems. Tortilla takes advantage of Attila's features for generating the geometric mesh, cross section library and source definitions. Tortilla can also read importance functions (like adjoint scalar flux) generated from deterministic calculations performed in Attila and use them to employ variance reduction schemes in the Monte Carlo simulation. The variance reduction techniques that are implemented in Tortilla are based on the CADIS (Consistent Adjoint Driven Importance Sampling) method and the LIFT (Local Importance Function Transform) method. These methods make use of the results from an adjoint deterministic calculation to bias the particle transport using techniques like source biasing, survival biasing, transport biasing and weight windows. The results obtained so far and the challenges faced in implementing the variance reduction techniques are reported here. (authors)
A user`s manual for MASH 1.0: A Monte Carlo Adjoint Shielding Code System
Johnson, J.O.
1992-03-01
The Monte Carlo Adjoint Shielding Code System, MASH, calculates neutron and gamma-ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air-over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system include the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. MASH is the successor to the Vehicle Code System (VCS) initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the ``dose importance`` of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response a a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user`s manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem (input data and selected output edits) for each code.
NASA Astrophysics Data System (ADS)
Franke, Brian C.; Kensek, Ronald P.; Prinja, Anil K.
2014-06-01
Stochastic-media simulations require numerous boundary crossings. We consider two Monte Carlo electron transport approaches and evaluate accuracy with numerous material boundaries. In the condensed-history method, approximations are made based on infinite-medium solutions for multiple scattering over some track length. Typically, further approximations are employed for material-boundary crossings where infinite-medium solutions become invalid. We have previously explored an alternative "condensed transport" formulation, a Generalized Boltzmann-Fokker-Planck GBFP method, which requires no special boundary treatment but instead uses approximations to the electron-scattering cross sections. Some limited capabilities for analog transport and a GBFP method have been implemented in the Integrated Tiger Series (ITS) codes. Improvements have been made to the condensed history algorithm. The performance of the ITS condensed-history and condensed-transport algorithms are assessed for material-boundary crossings. These assessments are made both by introducing artificial material boundaries and by comparison to analog Monte Carlo simulations.
Monte Carlo N-Particle Transport Code System To Simulate Time-Analysis Quantities.
PADOVANI, ENRICO
2012-04-15
Version: 00 US DOE 10CFR810 Jurisdiction. The Monte Carlo simulation of correlation measurements that rely on the detection of fast neutrons and photons from fission requires that particle emissions and interactions following a fission event be described as close to reality as possible. The -PoliMi extension to MCNP and to MCNPX was developed to simulate correlated-particle and the subsequent interactions as close as possible to the physical behavior. Initially, MCNP-PoliMi, a modification of MCNP4C, was developed. The first version was developed in 2001-2002 and released in early 2004 to the Radiation Safety Information Computational Center (RSICC). It was developed for research purposes, to simulate correlated counts in organic scintillation detectors, sensitive to fast neutrons and gamma rays. Originally, the field of application was nuclear safeguards; however subsequent improvements have enhanced the ability to model measurements in other research fields as well. During 2010-2011 the -PoliMi modification was ported into MCNPX-2.7.0, leading to the development of MCNPX-PoliMi. Now the -PoliMi v2.0 modifications are distributed as a patch to MCNPX-2.7.0 which currently is distributed in the RSICC PACKAGE BCC-004 MCNP6_BETA2/MCNP5/MCNPX. Also included in the package is MPPost, a versatile code that provides simulated detector response. By taking advantage of the modifications in MCNPX-PoliMi, MPPost can provide an accurate simulation of the detector response for a variety of detection scenarios.
Update on the Status of the FLUKA Monte Carlo Transport Code
NASA Technical Reports Server (NTRS)
Pinsky, L.; Anderson, V.; Empl, A.; Lee, K.; Smirnov, G.; Zapp, N; Ferrari, A.; Tsoulou, K.; Roesler, S.; Vlachoudis, V.; Battisoni, G.; Ceruti, F.; Gadioli, M. V.; Garzelli, M.; Muraro, S.; Rancati, T.; Sala, P.; Ballarini, R.; Ottolenghi, A.; Parini, V.; Scannicchio, D.; Pelliccioni, M.; Wilson, T. L.
2004-01-01
The FLUKA Monte Carlo transport code is a well-known simulation tool in High Energy Physics. FLUKA is a dynamic tool in the sense that it is being continually updated and improved by the authors. Here we review the progresses achieved in the last year on the physics models. From the point of view of hadronic physics, most of the effort is still in the field of nucleus--nucleus interactions. The currently available version of FLUKA already includes the internal capability to simulate inelastic nuclear interactions beginning with lab kinetic energies of 100 MeV/A up the the highest accessible energies by means of the DPMJET-II.5 event generator to handle the interactions for greater than 5 GeV/A and rQMD for energies below that. The new developments concern, at high energy, the embedding of the DPMJET-III generator, which represent a major change with respect to the DPMJET-II structure. This will also allow to achieve a better consistency between the nucleus-nucleus section with the original FLUKA model for hadron-nucleus collisions. Work is also in progress to implement a third event generator model based on the Master Boltzmann Equation approach, in order to extend the energy capability from 100 MeV/A down to the threshold for these reactions. In addition to these extended physics capabilities, structural changes to the programs input and scoring capabilities are continually being upgraded. In particular we want to mention the upgrades in the geometry packages, now capable of reaching higher levels of abstraction. Work is also proceeding to provide direct import into ROOT of the FLUKA output files for analysis and to deploy a user-friendly GUI input interface.
Monte Carlo N-Particle Transport Code System To Simulate Time-Analysis Quantities.
Energy Science and Technology Software Center (ESTSC)
2012-04-15
Version: 00 US DOE 10CFR810 Jurisdiction. The Monte Carlo simulation of correlation measurements that rely on the detection of fast neutrons and photons from fission requires that particle emissions and interactions following a fission event be described as close to reality as possible. The -PoliMi extension to MCNP and to MCNPX was developed to simulate correlated-particle and the subsequent interactions as close as possible to the physical behavior. Initially, MCNP-PoliMi, a modification of MCNP4C, wasmore » developed. The first version was developed in 2001-2002 and released in early 2004 to the Radiation Safety Information Computational Center (RSICC). It was developed for research purposes, to simulate correlated counts in organic scintillation detectors, sensitive to fast neutrons and gamma rays. Originally, the field of application was nuclear safeguards; however subsequent improvements have enhanced the ability to model measurements in other research fields as well. During 2010-2011 the -PoliMi modification was ported into MCNPX-2.7.0, leading to the development of MCNPX-PoliMi. Now the -PoliMi v2.0 modifications are distributed as a patch to MCNPX-2.7.0 which currently is distributed in the RSICC PACKAGE BCC-004 MCNP6_BETA2/MCNP5/MCNPX. Also included in the package is MPPost, a versatile code that provides simulated detector response. By taking advantage of the modifications in MCNPX-PoliMi, MPPost can provide an accurate simulation of the detector response for a variety of detection scenarios.« less
Development of a GPU-based Monte Carlo dose calculation code for coupled electron-photon transport.
Jia, Xun; Gu, Xuejun; Sempau, Josep; Choi, Dongju; Majumdar, Amitava; Jiang, Steve B
2010-06-01
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 2000 Phys. Med. Biol. 45 2263-91) on the GPU architecture under the CUDA platform. The implementation has been tested with respect to the original sequential DPM code on the CPU in phantoms with water-lung-water or water-bone-water slab geometry. A 20 MeV mono-energetic electron point source or a 6 MV photon point source is used in our validation. The results demonstrate adequate accuracy of our GPU implementation for both electron and photon beams in the radiotherapy energy range. Speed-up factors of about 5.0-6.6 times have been observed, using an NVIDIA Tesla C1060 GPU card against a 2.27 GHz Intel Xeon CPU processor. PMID:20463376
Burn, K W; Daffara, C; Gualdrini, G; Pierantoni, M; Ferrari, P
2007-01-01
The question of Monte Carlo simulation of radiation transport in voxel geometries is addressed. Patched versions of the MCNP and MCNPX codes are developed aimed at transporting radiation both in the standard geometry mode and in the voxel geometry treatment. The patched code reads an unformatted FORTRAN file derived from DICOM format data and uses special subroutines to handle voxel-to-voxel radiation transport. The various phases of the development of the methodology are discussed together with the new input options. Examples are given of employment of the code in internal and external dosimetry and comparisons with results from other groups are reported. PMID:17038404
Performance analysis of the Monte Carlo code MCNP4A for photon-based radiotherapy applications
DeMarco, J.J.; Solberg, T.D.; Wallace, R.E.; Smathers, J.B.
1995-12-31
The Los Alamos code MCNP4A (Monte Carlo M-Particle version 4A) is currently used to simulate a variety of problems ranging from nuclear reactor analysis to boron neutron capture therapy. This study is designed to evaluate MCNP4A as the dose calculation system for photon-based radiotherapy applications. A graphical user interface (MCNP Radiation Therapy) has been developed which automatically sets up the geometry and photon source requirements for three-dimensional simulations using Computed Tomography (CT) data. Preliminary results suggest the code is capable of calculating satisfactory dose distributions in a variety of simulated homogeneous and heterogeneous phantoms. The major drawback for this dosimetry system is the amount of time to obtain a statistically significant answer. MCNPRT allows the user to analyze the performance of MCNP4A as a function of material, geometry resolution and MCNP4A photon and electron physics parameters. A typical simulation geometry consists of a 10 MV photon point source incident on a 15 x 15 x 15 cm{sup 3} phantom composed of water voxels ranging in size from 10 x 10 x 10 mm{sup 3} to 2 x 2 x 2 mm{sup 3}. As the voxel size is decreased, a larger percentage of time is spent tracking photons through the voxelized geometry as opposed to the secondary electrons. A PRPR Patch file is under development that will optimize photon transport within the simulation phantom specifically for radiotherapy applications. MCNP4A also supports parallel processing capabilities via the Parallel Virtual Machine (PVM) message passing system. A dedicated network of five SUN SPARC2 processors produced a wall-clock speedup of 4.4 based on a simulation phantom containing 5 x 5 x 5 mm{sup 3} water voxels. The code was also tested on the 80 node IBM RS/6000 cluster at the Maui High Performance Computing Center (NHPCC). A non-dedicated system of 75 processors produces a wall clock speedup of 29 relative to one SUN SPARC2 computer.
Parodi, K; Ferrari, A; Sommerer, F; Paganetti, H
2008-01-01
Clinical investigations on post-irradiation PET/CT (positron emission tomography / computed tomography) imaging for in-vivo verification of treatment delivery and, in particular, beam range in proton therapy are underway at Massachusetts General Hospital (MGH). Within this project we have developed a Monte Carlo framework for CT-based calculation of dose and irradiation induced positron emitter distributions. Initial proton beam information is provided by a separate Geant4 Monte Carlo simulation modeling the treatment head. Particle transport in the patient is performed in the CT voxel geometry using the FLUKA Monte Carlo code. The implementation uses a discrete number of different tissue types with composition and mean density deduced from the CT scan. Scaling factors are introduced to account for the continuous Hounsfield Unit dependence of the mass density and of the relative stopping power ratio to water used by the treatment planning system (XiO (Computerized Medical Systems Inc.)). Resulting Monte Carlo dose distributions are generally found in good correspondence with calculations of the treatment planning program, except few cases (e.g. in the presence of air/tissue interfaces). Whereas dose is computed using standard FLUKA utilities, positron emitter distributions are calculated by internally combining proton fluence with experimental and evaluated cross-sections yielding 11C, 15O, 14O, 13N, 38K and 30P. Simulated positron emitter distributions yield PET images in good agreement with measurements. In this paper we describe in detail the specific implementation of the FLUKA calculation framework, which may be easily adapted to handle arbitrary phase spaces of proton beams delivered by other facilities or include more reaction channels based on additional cross-section data. Further, we demonstrate the effects of different acquisition time regimes (e.g., PET imaging during or after irradiation) on the intensity and spatial distribution of the irradiation
NASA Astrophysics Data System (ADS)
Parodi, K.; Ferrari, A.; Sommerer, F.; Paganetti, H.
2007-07-01
Clinical investigations on post-irradiation PET/CT (positron emission tomography/computed tomography) imaging for in vivo verification of treatment delivery and, in particular, beam range in proton therapy are underway at Massachusetts General Hospital (MGH). Within this project, we have developed a Monte Carlo framework for CT-based calculation of dose and irradiation-induced positron emitter distributions. Initial proton beam information is provided by a separate Geant4 Monte Carlo simulation modelling the treatment head. Particle transport in the patient is performed in the CT voxel geometry using the FLUKA Monte Carlo code. The implementation uses a discrete number of different tissue types with composition and mean density deduced from the CT scan. Scaling factors are introduced to account for the continuous Hounsfield unit dependence of the mass density and of the relative stopping power ratio to water used by the treatment planning system (XiO (Computerized Medical Systems Inc.)). Resulting Monte Carlo dose distributions are generally found in good correspondence with calculations of the treatment planning program, except a few cases (e.g. in the presence of air/tissue interfaces). Whereas dose is computed using standard FLUKA utilities, positron emitter distributions are calculated by internally combining proton fluence with experimental and evaluated cross-sections yielding 11C, 15O, 14O, 13N, 38K and 30P. Simulated positron emitter distributions yield PET images in good agreement with measurements. In this paper, we describe in detail the specific implementation of the FLUKA calculation framework, which may be easily adapted to handle arbitrary phase spaces of proton beams delivered by other facilities or include more reaction channels based on additional cross-section data. Further, we demonstrate the effects of different acquisition time regimes (e.g., PET imaging during or after irradiation) on the intensity and spatial distribution of the irradiation
Update On the Status of the FLUKA Monte Carlo Transport Code*
NASA Technical Reports Server (NTRS)
Ferrari, A.; Lorenzo-Sentis, M.; Roesler, S.; Smirnov, G.; Sommerer, F.; Theis, C.; Vlachoudis, V.; Carboni, M.; Mostacci, A.; Pelliccioni, M.
2006-01-01
The FLUKA Monte Carlo transport code is a well-known simulation tool in High Energy Physics. FLUKA is a dynamic tool in the sense that it is being continually updated and improved by the authors. We review the progress achieved since the last CHEP Conference on the physics models, some technical improvements to the code and some recent applications. From the point of view of the physics, improvements have been made with the extension of PEANUT to higher energies for p, n, pi, pbar/nbar and for nbars down to the lowest energies, the addition of the online capability to evolve radioactive products and get subsequent dose rates, upgrading of the treatment of EM interactions with the elimination of the need to separately prepare preprocessed files. A new coherent photon scattering model, an updated treatment of the photo-electric effect, an improved pair production model, new photon cross sections from the LLNL Cullen database have been implemented. In the field of nucleus-- nucleus interactions the electromagnetic dissociation of heavy ions has been added along with the extension of the interaction models for some nuclide pairs to energies below 100 MeV/A using the BME approach, as well as the development of an improved QMD model for intermediate energies. Both DPMJET 2.53 and 3 remain available along with rQMD 2.4 for heavy ion interactions above 100 MeV/A. Technical improvements include the ability to use parentheses in setting up the combinatorial geometry, the introduction of pre-processor directives in the input stream. a new random number generator with full 64 bit randomness, new routines for mathematical special functions (adapted from SLATEC). Finally, work is progressing on the deployment of a user-friendly GUI input interface as well as a CAD-like geometry creation and visualization tool. On the application front, FLUKA has been used to extensively evaluate the potential space radiation effects on astronauts for future deep space missions, the activation
Wu, Yunzhao; Tang, Zesheng
2014-01-01
In this paper, we model the reflectance of the lunar regolith by a new method combining Monte Carlo ray tracing and Hapke's model. The existing modeling methods exploit either a radiative transfer model or a geometric optical model. However, the measured data from an Interference Imaging spectrometer (IIM) on an orbiter were affected not only by the composition of minerals but also by the environmental factors. These factors cannot be well addressed by a single model alone. Our method implemented Monte Carlo ray tracing for simulating the large-scale effects such as the reflection of topography of the lunar soil and Hapke's model for calculating the reflection intensity of the internal scattering effects of particles of the lunar soil. Therefore, both the large-scale and microscale effects are considered in our method, providing a more accurate modeling of the reflectance of the lunar regolith. Simulation results using the Lunar Soil Characterization Consortium (LSCC) data and Chang'E-1 elevation map show that our method is effective and useful. We have also applied our method to Chang'E-1 IIM data for removing the influence of lunar topography to the reflectance of the lunar soil and to generate more realistic visualizations of the lunar surface. PMID:24526892
Wong, Un-Hong; Wu, Yunzhao; Wong, Hon-Cheng; Liang, Yanyan; Tang, Zesheng
2014-01-01
In this paper, we model the reflectance of the lunar regolith by a new method combining Monte Carlo ray tracing and Hapke's model. The existing modeling methods exploit either a radiative transfer model or a geometric optical model. However, the measured data from an Interference Imaging spectrometer (IIM) on an orbiter were affected not only by the composition of minerals but also by the environmental factors. These factors cannot be well addressed by a single model alone. Our method implemented Monte Carlo ray tracing for simulating the large-scale effects such as the reflection of topography of the lunar soil and Hapke's model for calculating the reflection intensity of the internal scattering effects of particles of the lunar soil. Therefore, both the large-scale and microscale effects are considered in our method, providing a more accurate modeling of the reflectance of the lunar regolith. Simulation results using the Lunar Soil Characterization Consortium (LSCC) data and Chang'E-1 elevation map show that our method is effective and useful. We have also applied our method to Chang'E-1 IIM data for removing the influence of lunar topography to the reflectance of the lunar soil and to generate more realistic visualizations of the lunar surface. PMID:24526892
Monte-Carlo Impurity transport simulations in the edge of the DIII-D tokamak using the MCI code
Evans, T.E.; Mahdavi, M.A.; Sager, G.T.; West, W.P.; Fenstermacher, M.E.; Meyer, W.H.; Porter, G.D.
1995-07-01
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.
Qin, Z.; Shoesmith, D.W.
2007-07-01
Based on a probabilistic model previously proposed, a Monte Carlo simulation code (EBSPA) has been developed to predict the lifetime of the engineered barriers system within the Yucca Mountain nuclear waste repository. The degradation modes considered in the EBSPA are general passive corrosion and hydrogen-induced cracking for the drip shield; and general passive corrosion, crevice corrosion and stress corrosion cracking for the waste package. Two scenarios have been simulated using the EBSPA code: (a) a conservative scenario for the conditions thought likely to prevail in the repository, and (b) an aggressive scenario in which the impact of the degradation processes is overstated. (authors)
DgSMC-B code: A robust and autonomous direct simulation Monte Carlo code for arbitrary geometries
NASA Astrophysics Data System (ADS)
Kargaran, H.; Minuchehr, A.; Zolfaghari, A.
2016-07-01
In this paper, we describe the structure of a new Direct Simulation Monte Carlo (DSMC) code that takes advantage of combinatorial geometry (CG) to simulate any rarefied gas flows Medias. The developed code, called DgSMC-B, has been written in FORTRAN90 language with capability of parallel processing using OpenMP framework. The DgSMC-B is capable of handling 3-dimensional (3D) geometries, which is created with first-and second-order surfaces. It performs independent particle tracking for the complex geometry without the intervention of mesh. In addition, it resolves the computational domain boundary and volume computing in border grids using hexahedral mesh. The developed code is robust and self-governing code, which does not use any separate code such as mesh generators. The results of six test cases have been presented to indicate its ability to deal with wide range of benchmark problems with sophisticated geometries such as airfoil NACA 0012. The DgSMC-B code demonstrates its performance and accuracy in a variety of problems. The results are found to be in good agreement with references and experimental data.
NASA Astrophysics Data System (ADS)
Kunz, Lothar; Kuhn, Frank M.; Deutschmann, Olaf
2015-07-01
So far most kinetic Monte Carlo (kMC) simulations of heterogeneously catalyzed gas phase reactions were limited to flat crystal surfaces. The newly developed program MoCKA (Monte Carlo Karlsruhe) combines graph-theoretical and lattice-based principles to be able to efficiently handle multiple lattices with a large number of sites, which account for different facets of the catalytic nanoparticle and the support material, and pursues a general approach, which is not restricted to a specific surface or reaction. The implementation uses the efficient variable step size method and applies a fast update algorithm for its process list. It is shown that the analysis of communication between facets and of (reverse) spillover effects is possible by rewinding the kMC simulation. Hence, this approach offers a wide range of new applications for kMC simulations in heterogeneous catalysis.
Monte Carlo Simulation of Siemens ONCOR Linear Accelerator with BEAMnrc and DOSXYZnrc Code
Jabbari, Keyvan; Anvar, Hossein Saberi; Tavakoli, Mohammad Bagher; Amouheidari, Alireza
2013-01-01
The Monte Carlo method is the most accurate method for simulation of radiation therapy equipment. The linear accelerators (linac) are currently the most widely used machines in radiation therapy centers. In this work, a Monte Carlo modeling of the Siemens ONCOR linear accelerator in 6 MV and 18 MV beams was performed. The results of simulation were validated by measurements in water by ionization chamber and extended dose range (EDR2) film in solid water. The linac's X-ray particular are so sensitive to the properties of primary electron beam. Square field size of 10 cm × 10 cm produced by the jaws was compared with ionization chamber and film measurements. Head simulation was performed with BEAMnrc and dose calculation with DOSXYZnrc for film measurements and 3ddose file produced by DOSXYZnrc analyzed used homemade MATLAB program. At 6 MV, the agreement between dose calculated by Monte Carlo modeling and direct measurement was obtained to the least restrictive of 1%, even in the build-up region. At 18 MV, the agreement was obtained 1%, except for in the build-up region. In the build-up region, the difference was 1% at 6 MV and 2% at 18 MV. The mean difference between measurements and Monte Carlo simulation is very small in both of ONCOR X-ray energy. The results are highly accurate and can be used for many applications such as patient dose calculation in treatment planning and in studies that model this linac with small field size like intensity-modulated radiation therapy technique. PMID:24672765
NASA Astrophysics Data System (ADS)
Lin, Yi-Chun; Liu, Yuan-Hao; Nievaart, Sander; Chen, Yen-Fu; Wu, Shu-Wei; Chou, Wen-Tsae; Jiang, Shiang-Huei
2011-10-01
High energy photon (over 10 MeV) and neutron beams adopted in radiobiology and radiotherapy always produce mixed neutron/gamma-ray fields. The Mg(Ar) ionization chambers are commonly applied to determine the gamma-ray dose because of its neutron insensitive characteristic. Nowadays, many perturbation corrections for accurate dose estimation and lots of treatment planning systems are based on Monte Carlo technique. The Monte Carlo codes EGSnrc, FLUKA, GEANT4, MCNP5, and MCNPX were used to evaluate energy dependent response functions of the Exradin M2 Mg(Ar) ionization chamber to a parallel photon beam with mono-energies from 20 keV to 20 MeV. For the sake of validation, measurements were carefully performed in well-defined (a) primary M-100 X-ray calibration field, (b) primary 60Co calibration beam, (c) 6-MV, and (d) 10-MV therapeutic beams in hospital. At energy region below 100 keV, MCNP5 and MCNPX both had lower responses than other codes. For energies above 1 MeV, the MCNP ITS-mode greatly resembled other three codes and the differences were within 5%. Comparing to the measured currents, MCNP5 and MCNPX using ITS-mode had perfect agreement with the 60Co, and 10-MV beams. But at X-ray energy region, the derivations reached 17%. This work shows us a better insight into the performance of different Monte Carlo codes in photon-electron transport calculation. Regarding the application of the mixed field dosimetry like BNCT, MCNP with ITS-mode is recognized as the most suitable tool by this work.
Dieudonne, C.; Dumonteil, E.; Malvagi, F.; Diop, C. M.
2013-07-01
For several years, Monte Carlo burnup/depletion codes have appeared, which couple a Monte Carlo code to simulate the neutron transport to a deterministic method that computes the medium depletion due to the neutron flux. Solving Boltzmann and Bateman equations in such a way allows to track fine 3 dimensional effects and to get rid of multi-group hypotheses done by deterministic solvers. The counterpart is the prohibitive calculation time due to the time-expensive Monte Carlo solver called at each time step. Therefore, great improvements in term of calculation time could be expected if one could get rid of Monte Carlo transport sequences. For example, it may seem interesting to run an initial Monte Carlo simulation only once, for the first time/burnup step, and then to use the concentration perturbation capability of the Monte Carlo code to replace the other time/burnup steps (the different burnup steps are seen like perturbations of the concentrations of the initial burnup step). This paper presents some advantages and limitations of this technique and preliminary results in terms of speed up and figure of merit. Finally, we will detail different possible calculation scheme based on that method. (authors)
VESTA 2.1.5 - Monte Carlo Depletion Interface Code; AURORA 1.0.0 - Depletion Analysis Tool.
HAECK, WIM
2013-03-21
Version 01 RSICC is authorized to distribute VESTA 2.1.5 for research and education purposes only. Requesters from NEA Data Bank member countries are advised to order VESTA 2.1.5 from the NEA Data Bank. Non-commercial and non-profit users from other OECD member countries (specifically Canada and the United States) may order VESTA 2.1.5 from RSICC. Users from non-OECD member countries and all commercial requesters are advised to contact the IRSN. VESTA is a Monte Carlo depletion interface code that is currently under development at IRSN (France). From its inception, VESTA is intended to be a generic interface code so that it will ultimately be capable of using any Monte-Carlo code or depletion module and that can be completely tailored to the users needs on practically all aspects of the code. For the current version, VESTA allows for the use of any version of MCNP(X) as the transport module and ORIGEN 2.2 or the built in PHOENIX module as the depletion module. A short overview of the main features of this version of the code is detailed in the Abstract.
VESTA 2.1.5 - Monte Carlo Depletion Interface Code; AURORA 1.0.0 - Depletion Analysis Tool.
Energy Science and Technology Software Center (ESTSC)
2013-03-21
Version 01 RSICC is authorized to distribute VESTA 2.1.5 for research and education purposes only. Requesters from NEA Data Bank member countries are advised to order VESTA 2.1.5 from the NEA Data Bank. Non-commercial and non-profit users from other OECD member countries (specifically Canada and the United States) may order VESTA 2.1.5 from RSICC. Users from non-OECD member countries and all commercial requesters are advised to contact the IRSN. VESTA is a Monte Carlo depletionmore » interface code that is currently under development at IRSN (France). From its inception, VESTA is intended to be a generic interface code so that it will ultimately be capable of using any Monte-Carlo code or depletion module and that can be completely tailored to the users needs on practically all aspects of the code. For the current version, VESTA allows for the use of any version of MCNP(X) as the transport module and ORIGEN 2.2 or the built in PHOENIX module as the depletion module. A short overview of the main features of this version of the code is detailed in the Abstract.« less
The Monte Carlo SRNA-VOX code for 3D proton dose distribution in voxelized geometry using CT data
NASA Astrophysics Data System (ADS)
Ilic, Radovan D.; Spasic-Jokic, Vesna; Belicev, Petar; Dragovic, Milos
2005-03-01
This paper describes the application of the SRNA Monte Carlo package for proton transport simulations in complex geometry and different material compositions. The SRNA package was developed for 3D dose distribution calculation in proton therapy and dosimetry and it was based on the theory of multiple scattering. The decay of proton induced compound nuclei was simulated by the Russian MSDM model and our own using ICRU 63 data. The developed package consists of two codes: the SRNA-2KG, which simulates proton transport in combinatorial geometry and the SRNA-VOX, which uses the voxelized geometry using the CT data and conversion of the Hounsfield's data to tissue elemental composition. Transition probabilities for both codes are prepared by the SRNADAT code. The simulation of the proton beam characterization by multi-layer Faraday cup, spatial distribution of positron emitters obtained by the SRNA-2KG code and intercomparison of computational codes in radiation dosimetry, indicate immediate application of the Monte Carlo techniques in clinical practice. In this paper, we briefly present the physical model implemented in the SRNA package, the ISTAR proton dose planning software, as well as the results of the numerical experiments with proton beams to obtain 3D dose distribution in the eye and breast tumour.
The Monte Carlo SRNA-VOX code for 3D proton dose distribution in voxelized geometry using CT data.
Ilić, Radovan D; Spasić-Jokić, Vesna; Belicev, Petar; Dragović, Milos
2005-03-01
This paper describes the application of the SRNA Monte Carlo package for proton transport simulations in complex geometry and different material compositions. The SRNA package was developed for 3D dose distribution calculation in proton therapy and dosimetry and it was based on the theory of multiple scattering. The decay of proton induced compound nuclei was simulated by the Russian MSDM model and our own using ICRU 63 data. The developed package consists of two codes: the SRNA-2KG, which simulates proton transport in combinatorial geometry and the SRNA-VOX, which uses the voxelized geometry using the CT data and conversion of the Hounsfield's data to tissue elemental composition. Transition probabilities for both codes are prepared by the SRNADAT code. The simulation of the proton beam characterization by multi-layer Faraday cup, spatial distribution of positron emitters obtained by the SRNA-2KG code and intercomparison of computational codes in radiation dosimetry, indicate immediate application of the Monte Carlo techniques in clinical practice. In this paper, we briefly present the physical model implemented in the SRNA package, the ISTAR proton dose planning software, as well as the results of the numerical experiments with proton beams to obtain 3D dose distribution in the eye and breast tumour. PMID:15798273
Liu, T.; Du, X.; Ji, W.; Xu, X. G.
2013-07-01
This paper describes the development of a Graphics Processing Unit (GPU) accelerated Monte Carlo photon transport code, ARCHER{sub GPU}, to perform CT imaging dose calculations with good accuracy and performance. The code simulates interactions of photons with heterogeneous materials. It contains a detailed CT scanner model and a family of patient phantoms. Several techniques are used to optimize the code for the GPU architecture. In the accuracy and performance test, a 142 kg adult male phantom was selected, and the CT scan protocol involved a whole-body axial scan, 20-mm x-ray beam collimation, 120 kVp and a pitch of 1. A total of 9 x 108 photons were simulated and the absorbed doses to 28 radiosensitive organs/tissues were calculated. The average percentage difference of the results obtained by the general-purpose production code MCNPX and ARCHER{sub GPU} was found to be less than 0.38%, indicating an excellent agreement. The total computation time was found to be 8,689, 139 and 56 minutes for MCNPX, ARCHER{sub CPU} (6-core) and ARCHER{sub GPU}, respectively, indicating a decent speedup. Under a recent grant funding from the NIH, the project aims at developing a Monte Carlo code with the capability of sub-minute CT organ dose calculations. (authors)
Energy Science and Technology Software Center (ESTSC)
2010-10-20
The "Monte Carlo Benchmark" (MCB) is intended to model the computatiional performance of Monte Carlo algorithms on parallel architectures. It models the solution of a simple heuristic transport equation using a Monte Carlo technique. The MCB employs typical features of Monte Carlo algorithms such as particle creation, particle tracking, tallying particle information, and particle destruction. Particles are also traded among processors using MPI calls.
A highly optimized vectorized code for Monte Carlo simulations of Su(3) lattice gauge theories
NASA Astrophysics Data System (ADS)
Barkai, D.; Moriarty, K. J. M.; Rebbi, C.
1984-04-01
New methods are introduced for improving the performance of the vectorized Monte Carlo SU(3) lattice gauge theory algorithm using the CDC CYBER 205. Structure, algorithm and programming considerations are discussed. The performance achieved for a 16 4 lattice on a 2-pipe system may be phrased in terms of the link update time or overall MFLOPS rates. For 32-bit arithmetic it is 36.3 μs/link for 8 hits per iteration (40.9 μs for 10 hits) or 101.5 MFLOPS.
Ultrafast vectorized multispin coding algorithm for the Monte Carlo simulation of the 3D Ising model
NASA Astrophysics Data System (ADS)
Wansleben, Stephan
1987-02-01
A new Monte Carlo algorithm for the 3D Ising model and its implementation on a CDC CYBER 205 is presented. This approach is applicable to lattices with sizes between 3·3·3 and 192·192·192 with periodic boundary conditions, and is adjustable to various kinetic models. It simulates a canonical ensemble at given temperature generating a new random number for each spin flip. For the Metropolis transition probability the speed is 27 ns per updates on a two-pipe CDC Cyber 205 with 2 million words physical memory, i.e. 1.35 times the cycle time per update or 38 million updates per second.
A highly optimized vectorized code for Monte Carlo simulations of SU(3) lattice gauge theories
NASA Technical Reports Server (NTRS)
Barkai, D.; Moriarty, K. J. M.; Rebbi, C.
1984-01-01
New methods are introduced for improving the performance of the vectorized Monte Carlo SU(3) lattice gauge theory algorithm using the CDC CYBER 205. Structure, algorithm and programming considerations are discussed. The performance achieved for a 16(4) lattice on a 2-pipe system may be phrased in terms of the link update time or overall MFLOPS rates. For 32-bit arithmetic, it is 36.3 microsecond/link for 8 hits per iteration (40.9 microsecond for 10 hits) or 101.5 MFLOPS.
Probability of initiation and extinction in the Mercury Monte Carlo code
McKinley, M. S.; Brantley, P. S.
2013-07-01
A Monte Carlo method for computing the probability of initiation has previously been implemented in Mercury. Recently, a new method based on the probability of extinction has been implemented as well. The methods have similarities from counting progeny to cycling in time, but they also have differences such as population control and statistical uncertainty reporting. The two methods agree very well for several test problems. Since each method has advantages and disadvantages, we currently recommend that both methods are used to compute the probability of criticality. (authors)
NASA Astrophysics Data System (ADS)
Tian, Zhen; Jiang Graves, Yan; Jia, Xun; Jiang, Steve B.
2014-10-01
Monte Carlo (MC) simulation is commonly considered as the most accurate method for radiation dose calculations. Commissioning of a beam model in the MC code against a clinical linear accelerator beam is of crucial importance for its clinical implementation. In this paper, we propose an automatic commissioning method for our GPU-based MC dose engine, gDPM. gDPM utilizes a beam model based on a concept of phase-space-let (PSL). A PSL contains a group of particles that are of the same type and close in space and energy. A set of generic PSLs was generated by splitting a reference phase-space file. Each PSL was associated with a weighting factor, and in dose calculations the particle carried a weight corresponding to the PSL where it was from. Dose for each PSL in water was pre-computed, and hence the dose in water for a whole beam under a given set of PSL weighting factors was the weighted sum of the PSL doses. At the commissioning stage, an optimization problem was solved to adjust the PSL weights in order to minimize the difference between the calculated dose and measured one. Symmetry and smoothness regularizations were utilized to uniquely determine the solution. An augmented Lagrangian method was employed to solve the optimization problem. To validate our method, a phase-space file of a Varian TrueBeam 6 MV beam was used to generate the PSLs for 6 MV beams. In a simulation study, we commissioned a Siemens 6 MV beam on which a set of field-dependent phase-space files was available. The dose data of this desired beam for different open fields and a small off-axis open field were obtained by calculating doses using these phase-space files. The 3D γ-index test passing rate within the regions with dose above 10% of dmax dose for those open fields tested was improved averagely from 70.56 to 99.36% for 2%/2 mm criteria and from 32.22 to 89.65% for 1%/1 mm criteria. We also tested our commissioning method on a six-field head-and-neck cancer IMRT plan. The
COOL: A code for Dynamic Monte Carlo Simulation of molecular dynamics
NASA Astrophysics Data System (ADS)
Barletta, Paolo
2012-02-01
Cool is a program to simulate evaporative and sympathetic cooling for a mixture of two gases co-trapped in an harmonic potential. The collisions involved are assumed to be exclusively elastic, and losses are due to evaporation from the trap. Each particle is followed individually in its trajectory, consequently properties such as spatial densities or energy distributions can be readily evaluated. The code can be used sequentially, by employing one output as input for another run. The code can be easily generalised to describe more complicated processes, such as the inclusion of inelastic collisions, or the possible presence of more than two species in the trap. New version program summaryProgram title: COOL Catalogue identifier: AEHJ_v2_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEHJ_v2_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 1 097 733 No. of bytes in distributed program, including test data, etc.: 18 425 722 Distribution format: tar.gz Programming language: C++ Computer: Desktop Operating system: Linux RAM: 500 Mbytes Classification: 16.7, 23 Catalogue identifier of previous version: AEHJ_v1_0 Journal reference of previous version: Comput. Phys. Comm. 182 (2011) 388 Does the new version supersede the previous version?: Yes Nature of problem: Simulation of the sympathetic process occurring for two molecular gases co-trapped in a deep optical trap. Solution method: The Direct Simulation Monte Carlo method exploits the decoupling, over a short time period, of the inter-particle interaction from the trapping potential. The particle dynamics is thus exclusively driven by the external optical field. The rare inter-particle collisions are considered with an acceptance/rejection mechanism, that is, by comparing a random number to the collisional probability
Basic physical and chemical information needed for development of Monte Carlo codes
Inokuti, M.
1993-08-01
It is important to view track structure analysis as an application of a branch of theoretical physics (i.e., statistical physics and physical kinetics in the language of the Landau school). Monte Carlo methods and transport equation methods represent two major approaches. In either approach, it is of paramount importance to use as input the cross section data that best represent the elementary microscopic processes. Transport analysis based on unrealistic input data must be viewed with caution, because results can be misleading. Work toward establishing the cross section data, which demands a wide scope of knowledge and expertise, is being carried out through extensive international collaborations. In track structure analysis for radiation biology, the need for cross sections for the interactions of electrons with DNA and neighboring protein molecules seems to be especially urgent. Finally, it is important to interpret results of Monte Carlo calculations fully and adequately. To this end, workers should document input data as thoroughly as possible and report their results in detail in many ways. Workers in analytic transport theory are then likely to contribute to the interpretation of the results.
Comparative Dosimetric Estimates of a 25 keV Electron Micro-beam with three Monte Carlo Codes
Mainardi, Enrico; Donahue, Richard J.; Blakely, Eleanor A.
2002-09-11
The calculations presented compare the different performances of the three Monte Carlo codes PENELOPE-1999, MCNP-4C and PITS, for the evaluation of Dose profiles from a 25 keV electron micro-beam traversing individual cells. The overall model of a cell is a water cylinder equivalent for the three codes but with a different internal scoring geometry: hollow cylinders for PENELOPE and MCNP, whereas spheres are used for the PITS code. A cylindrical cell geometry with scoring volumes with the shape of hollow cylinders was initially selected for PENELOPE and MCNP because of its superior simulation of the actual shape and dimensions of a cell and for its improved computer-time efficiency if compared to spherical internal volumes. Some of the transfer points and energy transfer that constitute a radiation track may actually fall in the space between spheres, that would be outside the spherical scoring volume. This internal geometry, along with the PENELOPE algorithm, drastically reduced the computer time when using this code if comparing with event-by-event Monte Carlo codes like PITS. This preliminary work has been important to address dosimetric estimates at low electron energies. It demonstrates that codes like PENELOPE can be used for Dose evaluation, even with such small geometries and energies involved, which are far below the normal use for which the code was created. Further work (initiated in Summer 2002) is still needed however, to create a user-code for PENELOPE that allows uniform comparison of exact cell geometries, integral volumes and also microdosimetric scoring quantities, a field where track-structure codes like PITS, written for this purpose, are believed to be superior.
Dewaraja, Yuni K.; Ljungberg, Michael; Majumdar, Amitava; Bose, Abhijit; Koral, Kenneth F.
2009-01-01
This paper reports the implementation of the SIMIND Monte Carlo code on an IBM SP2 distributed memory parallel computer. Basic aspects of running Monte Carlo particle transport calculations on parallel architectures are described. Our parallelization is based on equally partitioning photons among the processors and uses the Message Passing Interface (MPI) library for interprocessor communication and the Scalable Parallel Random Number Generator (SPRNG) to generate uncorrelated random number streams. These parallelization techniques are also applicable to other distributed memory architectures. A linear increase in computing speed with the number of processors is demonstrated for up to 32 processors. This speed-up is especially significant in Single Photon Emission Computed Tomography (SPECT) simulations involving higher energy photon emitters, where explicit modeling of the phantom and collimator is required. For 131I, the accuracy of the parallel code is demonstrated by comparing simulated and experimental SPECT images from a heart/thorax phantom. Clinically realistic SPECT simulations using the voxel-man phantom are carried out to assess scatter and attenuation correction. PMID:11809318
Optimization of a photoneutron source based on 10 MeV electron beam using Geant4 Monte Carlo code
NASA Astrophysics Data System (ADS)
Askri, Boubaker
2015-10-01
Geant4 Monte Carlo code has been used to conceive and optimize a simple and compact neutron source based on a 10 MeV electron beam impinging on a tungsten target adjoined to a beryllium target. For this purpose, a precise photonuclear reaction cross-section model issued from the International Atomic Energy Agency (IAEA) database was linked to Geant4 to accurately simulate the interaction of low energy bremsstrahlung photons with beryllium material. A benchmark test showed that a good agreement was achieved when comparing the emitted neutron flux spectra predicted by Geant4 and Fluka codes for a beryllium cylinder bombarded with a 5 MeV photon beam. The source optimization was achieved through a two stage Monte Carlo simulation. In the first stage, the distributions of the seven phase space coordinates of the bremsstrahlung photons at the boundaries of the tungsten target were determined. In the second stage events corresponding to photons emitted according to these distributions were tracked. A neutron yield of 4.8 × 1010 neutrons/mA/s was obtained at 20 cm from the beryllium target. A thermal neutron yield of 1.5 × 109 neutrons/mA/s was obtained after introducing a spherical shell of polyethylene as a neutron moderator.
Proton Upset Monte Carlo Simulation
NASA Technical Reports Server (NTRS)
O'Neill, Patrick M.; Kouba, Coy K.; Foster, Charles C.
2009-01-01
The Proton Upset Monte Carlo Simulation (PROPSET) program calculates the frequency of on-orbit upsets in computer chips (for given orbits such as Low Earth Orbit, Lunar Orbit, and the like) from proton bombardment based on the results of heavy ion testing alone. The software simulates the bombardment of modern microelectronic components (computer chips) with high-energy (.200 MeV) protons. The nuclear interaction of the proton with the silicon of the chip is modeled and nuclear fragments from this interaction are tracked using Monte Carlo techniques to produce statistically accurate predictions.
Khajeh, Masoud; Safigholi, Habib
2016-03-01
A miniature X-ray source has been optimized for electronic brachytherapy. The cooling fluid for this device is water. Unlike the radionuclide brachytherapy sources, this source is able to operate at variable voltages and currents to match the dose with the tumor depth. First, Monte Carlo (MC) optimization was performed on the tungsten target-buffer thickness layers versus energy such that the minimum X-ray attenuation occurred. Second optimization was done on the selection of the anode shape based on the Monte Carlo in water TG-43U1 anisotropy function. This optimization was carried out to get the dose anisotropy functions closer to unity at any angle from 0° to 170°. Three anode shapes including cylindrical, spherical, and conical were considered. Moreover, by Computational Fluid Dynamic (CFD) code the optimal target-buffer shape and different nozzle shapes for electronic brachytherapy were evaluated. The characterization criteria of the CFD were the minimum temperature on the anode shape, cooling water, and pressure loss from inlet to outlet. The optimal anode was conical in shape with a conical nozzle. Finally, the TG-43U1 parameters of the optimal source were compared with the literature. PMID:26966563
Khajeh, Masoud; Safigholi, Habib
2015-01-01
A miniature X-ray source has been optimized for electronic brachytherapy. The cooling fluid for this device is water. Unlike the radionuclide brachytherapy sources, this source is able to operate at variable voltages and currents to match the dose with the tumor depth. First, Monte Carlo (MC) optimization was performed on the tungsten target-buffer thickness layers versus energy such that the minimum X-ray attenuation occurred. Second optimization was done on the selection of the anode shape based on the Monte Carlo in water TG-43U1 anisotropy function. This optimization was carried out to get the dose anisotropy functions closer to unity at any angle from 0° to 170°. Three anode shapes including cylindrical, spherical, and conical were considered. Moreover, by Computational Fluid Dynamic (CFD) code the optimal target-buffer shape and different nozzle shapes for electronic brachytherapy were evaluated. The characterization criteria of the CFD were the minimum temperature on the anode shape, cooling water, and pressure loss from inlet to outlet. The optimal anode was conical in shape with a conical nozzle. Finally, the TG-43U1 parameters of the optimal source were compared with the literature. PMID:26966563
O'Brien, M J; Procassini, R J; Joy, K I
2009-03-09
Validation of the problem definition and analysis of the results (tallies) produced during a Monte Carlo particle transport calculation can be a complicated, time-intensive processes. The time required for a person to create an accurate, validated combinatorial geometry (CG) or mesh-based representation of a complex problem, free of common errors such as gaps and overlapping cells, can range from days to weeks. The ability to interrogate the internal structure of a complex, three-dimensional (3-D) geometry, prior to running the transport calculation, can improve the user's confidence in the validity of the problem definition. With regard to the analysis of results, the process of extracting tally data from printed tables within a file is laborious and not an intuitive approach to understanding the results. The ability to display tally information overlaid on top of the problem geometry can decrease the time required for analysis and increase the user's understanding of the results. To this end, our team has integrated VisIt, a parallel, production-quality visualization and data analysis tool into Mercury, a massively-parallel Monte Carlo particle transport code. VisIt provides an API for real time visualization of a simulation as it is running. The user may select which plots to display from the VisIt GUI, or by sending VisIt a Python script from Mercury. The frequency at which plots are updated can be set and the user can visualize the simulation results as it is running.
Towards scalable parellelism in Monte Carlo particle transport codes using remote memory access
Romano, Paul K; Brown, Forrest B; Forget, Benoit
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.
Creation of problem-dependent Doppler-broadened cross sections in the KENO Monte Carlo code
Hart, Shane W. D.; Celik, Cihangir; Maldonado, G. Ivan; Leal, Luiz C.
2015-11-06
In this paper, we introduce a quick method for improving the accuracy of Monte Carlo simulations by generating one- and two-dimensional cross sections at a user-defined temperature before performing transport calculations. A finite difference method is used to Doppler-broaden cross sections to the desired temperature, and unit-base interpolation is done to generate the probability distributions for double differential two-dimensional thermal moderator cross sections at any arbitrarily user-defined temperature. The accuracy of these methods is tested using a variety of contrived problems. In addition, various benchmarks at elevated temperatures are modeled, and results are compared with benchmark results. Lastly, the problem-dependent cross sections are observed to produce eigenvalue estimates that are closer to the benchmark results than those without the problem-dependent cross sections.
Creation of problem-dependent Doppler-broadened cross sections in the KENO Monte Carlo code
Hart, Shane W. D.; Celik, Cihangir; Maldonado, G. Ivan; Leal, Luiz C.
2015-11-06
In this paper, we introduce a quick method for improving the accuracy of Monte Carlo simulations by generating one- and two-dimensional cross sections at a user-defined temperature before performing transport calculations. A finite difference method is used to Doppler-broaden cross sections to the desired temperature, and unit-base interpolation is done to generate the probability distributions for double differential two-dimensional thermal moderator cross sections at any arbitrarily user-defined temperature. The accuracy of these methods is tested using a variety of contrived problems. In addition, various benchmarks at elevated temperatures are modeled, and results are compared with benchmark results. Lastly, the problem-dependentmore » cross sections are observed to produce eigenvalue estimates that are closer to the benchmark results than those without the problem-dependent cross sections.« less
Characterisation of the TRIUMF neutron facility using a Monte Carlo simulation code.
Monk, S D; Abram, T; Joyce, M J
2015-04-01
Here, the characterisation of the high-energy neutron field at TRIUMF (The Tri Universities Meson Facility, Vancouver, British Columbia) with Monte Carlo simulation software is described. The package used is MCNPX version 2.6.0, with the neutron fluence rate determined at three locations within the TRIUMF Thermal Neutron Facility (TNF), including the exit of the neutron channel where users of the facility can test devices that may be susceptible to the effects of this form of radiation. The facility is often used to roughly emulate the field likely to be encountered at high altitudes due to radiation of galactic origin and thus the simulated information is compared with the energy spectrum calculated to be due to neutron radiation of cosmic origin at typical aircraft altitudes. The calculated values were also compared with neutron flux measurements that were estimated using the activation of various foils by the staff of the facility, showing agreement within an order of magnitude. PMID:25342608
Code System to Perform Monte Carlo Simulation of Electron Gamma-Ray Showers in Arbitrary Marerials.
Energy Science and Technology Software Center (ESTSC)
2002-10-15
Version 00 PENELOPE performs Monte Carlo simulation of electron-photon showers in arbitrary materials. Initially, it was devised to simulate the PENetration and Energy LOss of Positrons and Electrons in matter; photons were introduced later. The adopted scattering model gives a reliable description of radiation transport in the energy range from a few hundred eV to about 1GeV. PENELOPE generates random electron-photon showers in complex material structures consisting of any number of distinct homogeneous regions (bodies)more » with different compositions. The Penelope Forum list archives and other information can be accessed at http://www.nea.fr/lists/penelope.html. PENELOPE-MPI extends capabilities of PENELOPE-2001 (RSICC C00682MNYCP02; NEA-1525/05) by providing for usage of MPI type parallel drivers and extends the original version's ability to read different types of input data sets such as voxel. The motivation is to increase efficiency of Monte Carlo simulations for medical applications. The physics of the calculations have not been changed, and the original description of PENELOPE-2001 (which follows) is still valid. PENELOPE-2001 contains substantial changes and improvements to the previous versions 1996 and 2000. As for the physics, the model for electron/positron elastic scattering has been revised. Bremsstrahlung emission is now simulated using partial-wave data instead of analytical approximate formulae. Photoelectric absorption in K and L-shells is described from the corresponding partial cross sections. Fluorescence radiation from vacancies in K and L-shells is followed. Refinements were also introduced in electron/positron transport mechanics, mostly to account for energy dependence of the mean free paths for hard events. Simulation routines were re-programmed in a more structured way, and new example MAIN programs were written with a more flexible input and expanded output.« less
A High-Accurate and High-Efficient Monte Carlo Code by Improved Molière Functions with Ionization
NASA Astrophysics Data System (ADS)
Nakatsuka, Takao; Okei, Kazuhide
2003-07-01
Although the Molière theory of multiple Coulomb scattering is less accue rate in tracing solid angles than the Goudsmit and Saunderson theory due to the small angle approximation, it still acts very important roles in developments of high-efficient simulation codes of relativistic charged particles like cosmic-ray particles. Molière expansion is well explained by the physical model, that is the e normal distribution attributing to the high-frequent moderate scatterings and subsequent correction terms attributing to the additive large-angle scatterings. Based on these physical concepts, we have improved a high-accurate and highefficient Monte Carlo code taking account of ionization loss.
Walsh, J. A.; Palmer, T. S.; Urbatsch, T. J.
2013-07-01
A new method for generating discrete scattering cross sections to be used in charged particle transport calculations is investigated. The method of data generation is presented and compared to current methods for obtaining discrete cross sections. The new, more generalized approach allows greater flexibility in choosing a cross section model from which to derive discrete values. Cross section data generated with the new method is verified through a comparison with discrete data obtained with an existing method. Additionally, a charged particle transport capability is demonstrated in the time-dependent Implicit Monte Carlo radiative transfer code package, Milagro. The implementation of this capability is verified using test problems with analytic solutions as well as a comparison of electron dose-depth profiles calculated with Milagro and an already-established electron transport code. An initial investigation of a preliminary integration of the discrete cross section generation method with the new charged particle transport capability in Milagro is also presented. (authors)
NASA Astrophysics Data System (ADS)
Gifford, Kent A.; Horton, John L., Jr.; Wareing, Todd A.; Failla, Gregory; Mourtada, Firas
2006-05-01
Radiotherapy calculations often involve complex geometries such as interfaces between materials of vastly differing atomic number, such as lung, bone and/or air interfaces. Monte Carlo methods have been used to calculate accurately the perturbation effects of the interfaces. However, these methods can be computationally expensive for routine clinical calculations. An alternative approach is to solve the Boltzmann equation deterministically. We present one such deterministic code, Attila™. Further, we computed a brachytherapy example and an external beam benchmark to compare the results with data previously calculated by MCNPX and EGS4. Our data suggest that the presented deterministic code is as accurate as EGS4 and MCNPX for the transport geometries examined in this study.
NASA Astrophysics Data System (ADS)
Lazarakis, P.; Bug, M. U.; Gargioni, E.; Guatelli, S.; Rabus, H.; Rosenfeld, A. B.
2012-03-01
The concept of nanodosimetry is based on the assumption that initial damage to cells is related to the number of ionizations (the ionization cluster size) directly produced by single particles within, or in the close vicinity of, short segments of DNA. The ionization cluster-size distribution and other nanodosimetric quantities, however, are not directly measurable in biological targets and our current knowledge is mostly based on numerical simulations of particle tracks in water, calculating track structure parameters for nanometric target volumes. The assessment of nanodosimetric quantities derived from particle-track calculations using different Monte Carlo codes plays, therefore, an important role for a more accurate evaluation of the initial damage to cells and, as a consequence, of the biological effectiveness of ionizing radiation. The aim of this work is to assess the differences in the calculated nanodosimetric quantities obtained with Geant4-DNA as compared to those of the ad hoc particle-track Monte Carlo code ‘PTra’ developed at Physikalisch-Technische Bundesanstalt (PTB), Germany. The comparison of the two codes was made for incident electrons of energy in the range between 50 eV and 10 keV, for protons of energy between 300 keV and 10 MeV, and for alpha particles of energy between 1 and 10 MeV as these were the energy ranges available in both codes at the time this investigation was carried out. Good agreement was found for nanodosimetric characteristics of track structure calculated in the high-energy range of each particle type. For lower energies, significant differences were observed, most notably in the estimates of the biological effectiveness. The largest relative differences obtained were over 50%; however, generally the order of magnitude was between 10% and 20%.
Lazarakis, P; Bug, M U; Gargioni, E; Guatelli, S; Rabus, H; Rosenfeld, A B
2012-03-01
The concept of nanodosimetry is based on the assumption that initial damage to cells is related to the number of ionizations (the ionization cluster size) directly produced by single particles within, or in the close vicinity of, short segments of DNA. The ionization cluster-size distribution and other nanodosimetric quantities, however, are not directly measurable in biological targets and our current knowledge is mostly based on numerical simulations of particle tracks in water, calculating track structure parameters for nanometric target volumes. The assessment of nanodosimetric quantities derived from particle-track calculations using different Monte Carlo codes plays, therefore, an important role for a more accurate evaluation of the initial damage to cells and, as a consequence, of the biological effectiveness of ionizing radiation. The aim of this work is to assess the differences in the calculated nanodosimetric quantities obtained with Geant4-DNA as compared to those of the ad hoc particle-track Monte Carlo code 'PTra' developed at Physikalisch-Technische Bundesanstalt (PTB), Germany. The comparison of the two codes was made for incident electrons of energy in the range between 50 eV and 10 keV, for protons of energy between 300 keV and 10 MeV, and for alpha particles of energy between 1 and 10 MeV as these were the energy ranges available in both codes at the time this investigation was carried out. Good agreement was found for nanodosimetric characteristics of track structure calculated in the high-energy range of each particle type. For lower energies, significant differences were observed, most notably in the estimates of the biological effectiveness. The largest relative differences obtained were over 50%; however, generally the order of magnitude was between 10% and 20%. PMID:22330641
Energy Science and Technology Software Center (ESTSC)
2006-05-09
The Monte Carlo example programs VARHATOM and DMCATOM are two small, simple FORTRAN programs that illustrate the use of the Monte Carlo Mathematical technique for calculating the ground state energy of the hydrogen atom.
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.
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.
Simulation of the full-core pin-model by JMCT Monte Carlo neutron-photon transport code
Li, D.; Li, G.; Zhang, B.; Shu, L.; Shangguan, D.; Ma, Y.; Hu, Z.
2013-07-01
Since the large numbers of cells over a million, the tallies over a hundred million and the particle histories over ten billion, the simulation of the full-core pin-by-pin model has become a real challenge for the computers and the computational methods. On the other hand, the basic memory of the model has exceeded the limit of a single CPU, so the spatial domain and data decomposition must be considered. JMCT (J Monte Carlo Transport code) has successful fulfilled the simulation of the full-core pin-by-pin model by the domain decomposition and the nested parallel computation. The k{sub eff} and flux of each cell are obtained. (authors)
NASA Astrophysics Data System (ADS)
Infantino, Angelo; Oehlke, Elisabeth; Mostacci, Domiziano; Schaffer, Paul; Trinczek, Michael; Hoehr, Cornelia
2016-01-01
The Monte Carlo code FLUKA is used to simulate the production of a number of positron emitting radionuclides, 18F, 13N, 94Tc, 44Sc, 68Ga, 86Y, 89Zr, 52Mn, 61Cu and 55Co, on a small medical cyclotron with a proton beam energy of 13 MeV. Experimental data collected at the TR13 cyclotron at TRIUMF agree within a factor of 0.6 ± 0.4 with the directly simulated data, except for the production of 55Co, where the simulation underestimates the experiment by a factor of 3.4 ± 0.4. The experimental data also agree within a factor of 0.8 ± 0.6 with the convolution of simulated proton fluence and cross sections from literature. Overall, this confirms the applicability of FLUKA to simulate radionuclide production at 13 MeV proton beam energy.
Evans, T.E.; Leonard, A.W.; West, W.P.; Finkenthal, D.F.; Fenstermacher, M.E.; Porter, G.D.
1998-08-01
Experimentally measured carbon line emissions and total radiated power distributions from the DIII-D divertor and Scrape-Off Layer (SOL) are compared to those calculated with the Monte Carlo Impurity (MCI) model. A UEDGE background plasma is used in MCI with the Roth and Garcia-Rosales (RG-R) chemical sputtering model and/or one of six physical sputtering models. While results from these simulations do not reproduce all of the features seen in the experimentally measured radiation patterns, the total radiated power calculated in MCI is in relatively good agreement with that measured by the DIII-D bolometric system when the Smith78 physical sputtering model is coupled to RG-R chemical sputtering in an unaltered UEDGE plasma. Alternatively, MCI simulations done with UEDGE background ion temperatures along the divertor target plates adjusted to better match those measured in the experiment resulted in three physical sputtering models which when coupled to the RG-R model gave a total radiated power that was within 10% of measured value.
PUVA: A Monte Carlo code for intra-articular PUVA treatment of arthritis
Descalle, M.A.; Laing, T.J.; Martin, W.R.
1996-12-31
Current rheumatoid arthritis treatments are only partially successful. Intra-articular psoralen-ultraviolet light (PUVA) phototherapy appears to be a new and valid alternative. Ultraviolet laser light (UVA) delivered in the knee joint through a fiber optic is used in combination with 8-methoxypsoralen (8-MOP), a light-sensitive chemical administered orally. A few hours after ingestion, the psoralen has diffused in all body cells. Once activated by UVA light, it binds to biological molecules, inhabiting cell division and ultimately causing local control of the arthritis. The magnitude of the response is proportional to the number of photoproducts delivered to tissues (i.e., the number of absorbed photons): the PUVA treatment will only be effective if a sufficient and relatively uniform dose is delivered to the diseased synovial tissues, while sparing other tissues such as cartilage. An application is being developed, based on analog Monte Carlo methods, to predict photon densities in tissues and the minimum number of intra-articular catheter positions necessary to ensure proper treatment of the diseased zone. Other interesting aspects of the problem deal with the compexity of the joint geometry, the physics of light scattering in tissues (a relatively new field of research that is not fully understood because of the variety of tissues and tissue components), and, finally, the need to include optic laws (reflection and refraction) at interfaces.
Monte Carlo simulation of a multi-leaf collimator design for telecobalt machine using BEAMnrc code
Ayyangar, Komanduri M.; Kumar, M. Dinesh; Narayan, Pradush; Jesuraj, Fenedit; Raju, M. R.
2010-01-01
This investigation aims to design a practical multi-leaf collimator (MLC) system for the cobalt teletherapy machine and check its radiation properties using the Monte Carlo (MC) method. The cobalt machine was modeled using the BEAMnrc Omega-Beam MC system, which could be freely downloaded from the website of the National Research Council (NRC), Canada. Comparison with standard depth dose data tables and the theoretically modeled beam showed good agreement within 2%. An MLC design with low melting point alloy (LMPA) was tested for leakage properties of leaves. The LMPA leaves with a width of 7 mm and height of 6 cm, with tongue and groove of size 2 mm wide by 4 cm height, produced only 4% extra leakage compared to 10 cm height tungsten leaves. With finite 60Co source size, the interleaf leakage was insignificant. This analysis helped to design a prototype MLC as an accessory mount on a cobalt machine. The complete details of the simulation process and analysis of results are discussed. PMID:20177567
Chin, P.W. . E-mail: mary.chin@physics.org
2005-10-15
This project developed a solution for verifying external photon beam radiotherapy. The solution is based on a calibration chain for deriving portal dose maps from acquired portal images, and a calculation framework for predicting portal dose maps. Quantitative comparison between acquired and predicted portal dose maps accomplishes both geometric (patient positioning with respect to the beam) and dosimetric (two-dimensional fluence distribution of the beam) verifications. A disagreement would indicate that beam delivery had not been according to plan. The solution addresses the clinical need for verifying radiotherapy both pretreatment (without the patient in the beam) and on treatment (with the patient in the beam). Medical linear accelerators mounted with electronic portal imaging devices (EPIDs) were used to acquire portal images. Two types of EPIDs were investigated: the amorphous silicon (a-Si) and the scanning liquid ion chamber (SLIC). The EGSnrc family of Monte Carlo codes were used to predict portal dose maps by computer simulation of radiation transport in the beam-phantom-EPID configuration. Monte Carlo simulations have been implemented on several levels of high throughput computing (HTC), including the grid, to reduce computation time. The solution has been tested across the entire clinical range of gantry angle, beam size (5 cmx5 cm to 20 cmx20 cm), and beam-patient and patient-EPID separations (4 to 38 cm). In these tests of known beam-phantom-EPID configurations, agreement between acquired and predicted portal dose profiles was consistently within 2% of the central axis value. This Monte Carlo portal dosimetry solution therefore achieved combined versatility, accuracy, and speed not readily achievable by other techniques.
Habib, B; Poumarede, B; Tola, F; Barthe, J
2010-01-01
The aim of the present study is to demonstrate the potential of accelerated dose calculations, using the fast Monte Carlo (MC) code referred to as PENFAST, rather than the conventional MC code PENELOPE, without losing accuracy in the computed dose. For this purpose, experimental measurements of dose distributions in homogeneous and inhomogeneous phantoms were compared with simulated results using both PENELOPE and PENFAST. The simulations and experiments were performed using a Saturne 43 linac operated at 12 MV (photons), and at 18 MeV (electrons). Pre-calculated phase space files (PSFs) were used as input data to both the PENELOPE and PENFAST dose simulations. Since depth-dose and dose profile comparisons between simulations and measurements in water were found to be in good agreement (within +/-1% to 1 mm), the PSF calculation is considered to have been validated. In addition, measured dose distributions were compared to simulated results in a set of clinically relevant, inhomogeneous phantoms, consisting of lung and bone heterogeneities in a water tank. In general, the PENFAST results agree to within a 1% to 1 mm difference with those produced by PENELOPE, and to within a 2% to 2 mm difference with measured values. Our study thus provides a pre-clinical validation of the PENFAST code. It also demonstrates that PENFAST provides accurate results for both photon and electron beams, equivalent to those obtained with PENELOPE. CPU time comparisons between both MC codes show that PENFAST is generally about 9-21 times faster than PENELOPE. PMID:19342258
NASA Astrophysics Data System (ADS)
Gudmundsson, J. T.; Lieberman, M. A.; Wang, Ying; Verboncoeur, J. P.
2009-10-01
The oopd1 particle-in-cell Monte Carlo (PIC-MC) code is used to simulate a capacitively coupled discharge in oxygen. oopd1 is a one-dimensional object-oriented PIC-MC code [1] in which the model system has one spatial dimension and three velocity components. It contains models for planar, cylindrical, and spherical geometries and replaces the XPDx1 series [2], which is not object-oriented. The revised oxygen model includes, in addition to electrons, the oxygen molecule in ground state, the oxygen atom in ground state, the negative ion O^-, and the positive ions O^+ and O2^+. The cross sections for the collisions among the oxygen species have been significantly revised from earlier work using the xpdp1 code [3]. Here we explore the electron energy distribution function (EEDF), the ion energy distribution function (IEDF) and the density profiles for various pressures and driving frequencies. In particular we investigate the influence of the O^+ ion on the IEDF, we explore the influence of multiple driving frequencies, and we do comparisons to the previous xpdx1 codes. [1] J. P. Verboncoeur, A. B. Langdon, and N. T. Gladd, Comp. Phys. Comm. 87 (1995) 199 [2] J. P. Verboncoeur, M. V. Alves, V. Vahedi, and C. K. Birdsall, J. Comp. Physics 104 (1993) 321 [2] V. Vahedi and M. Surendra, Comp. Phys. Comm. 87 (1995) 179
NASA Astrophysics Data System (ADS)
Pedrocchi, Fabio L.; Bonesteel, N. E.; DiVincenzo, David P.
2015-09-01
The Majorana code is an example of a stabilizer code where the quantum information is stored in a system supporting well-separated Majorana bound states (MBSs). We focus on one-dimensional realizations of the Majorana code, as well as networks of such structures, and investigate their lifetime when coupled to a parity-preserving thermal environment. We apply the Davies prescription, a standard method that describes the basic aspects of a thermal environment, and derive a master equation in the Born-Markov limit. We first focus on a single wire with immobile MBSs and perform error correction to annihilate thermal excitations. In the high-temperature limit, we show both analytically and numerically that the lifetime of the Majorana qubit grows logarithmically with the size of the wire. We then study a trijunction with four MBSs when braiding is executed. We study the occurrence of dangerous error processes that prevent the lifetime of the Majorana code from growing with the size of the trijunction. The origin of the dangerous processes is the braiding itself, which separates pairs of excitations and renders the noise nonlocal; these processes arise from the basic constraints of moving MBSs in one-dimensional (1D) structures. We confirm our predictions with Monte Carlo simulations in the low-temperature regime, i.e., the regime of practical relevance. Our results put a restriction on the degree of self-correction of this particular 1D topological quantum computing architecture.
NASA Astrophysics Data System (ADS)
Takeda, N.; Kudo, K.; Toyokawa, H.; Torii, T.; Hashimoto, M.; Sugita, T.; Dietze, G.; Yang, X.
1999-02-01
A Monte Carlo code Neutron RESPonse function for Gas counters (NRESPG) has been developed for the calculation of neutron response functions and efficiencies for neutron energies up to 20 MeV, which can be applied for 3He, H 2, or BF 3 gas proportional counters with or without moderator. This code can simulate the neutron behavior in a two-dimensional detector configuration and treat the thermal motion of a moderator atom which becomes important as the neutron energy becomes sufficiently low. Further, a more precise measured data was taken to simulate the position-dependent gas multiplication in the sensitive and insensitive gas region of a proportional counter. The NRESPG code has been applied for the calculation of response functions of 3He cylindrical proportional counters to determine neutron energy and neutron fluence in a monoenergetic calibration field. Thus, a remarkable discrepancy in the lower portion of the full-energy peak produced by the 3He(n,p)T reaction can be removed which results in a good agreement between simulations and experiments. The code has been also used for the simulation of the response of a McTaggart-type long counter consisting of a central cylindrical BF 3 counter surrounded by a polyethylene moderator. The results of the NRESPG simulations were compared with those obtained from MCNP calculations.
Botta, F.; Mairani, A.; Battistoni, G.; Cremonesi, M.; Di Dia, A.; Fasso, A.; Ferrari, A.; Ferrari, M.; Paganelli, G.; Pedroli, G.; Valente, M.
2011-07-15
Purpose: The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, fluka Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, fluka has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one. Methods: fluka DPKs have been calculated in both water and compact bone for monoenergetic electrons (10{sup -3} MeV) and for beta emitting isotopes commonly used for therapy ({sup 89}Sr, {sup 90}Y, {sup 131}I, {sup 153}Sm, {sup 177}Lu, {sup 186}Re, and {sup 188}Re). Point isotropic sources have been simulated at the center of a water (bone) sphere, and deposed energy has been tallied in concentric shells. fluka outcomes have been compared to penelope v.2008 results, calculated in this study as well. Moreover, in case of monoenergetic electrons in water, comparison with the data from the literature (etran, geant4, mcnpx) has been done. Maximum percentage differences within 0.8{center_dot}R{sub CSDA} and 0.9{center_dot}R{sub CSDA} for monoenergetic electrons (R{sub CSDA} being the continuous slowing down approximation range) and within 0.8{center_dot}X{sub 90} and 0.9{center_dot}X{sub 90} for isotopes (X{sub 90} being the radius of the sphere in which 90% of the emitted energy is absorbed) have been computed, together with the average percentage difference within 0.9{center_dot}R{sub CSDA} and 0.9{center_dot}X{sub 90} for electrons and isotopes, respectively. Results: Concerning monoenergetic electrons
Botta, F; Di Dia, A; Pedroli, G; Mairani, A; Battistoni, G; Fasso, A; Ferrari, A; Ferrari, M; Paganelli, G; Valente, M
2011-06-01
The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, fluka Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, fluka has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one.Methods: fluka DPKs have been calculated in both water and compact bone for monoenergetic electrons (10–3 MeV) and for beta emitting isotopes commonly used for therapy (89Sr, 90Y, 131I, 153Sm, 177Lu, 186Re, and 188Re). Point isotropic sources have been simulated at the center of a water (bone) sphere, and deposed energy has been tallied in concentric shells. fluka outcomes have been compared to penelope v.2008 results, calculated in this study as well. Moreover, in case of monoenergetic electrons in water, comparison with the data from the literature (etran, geant4, mcnpx) has been done. Maximum percentage differences within 0.8·RCSDA and 0.9·RCSDA for monoenergetic electrons (RCSDA being the continuous slowing down approximation range) and within 0.8·X90 and 0.9·X90 for isotopes (X90 being the radius of the sphere in which 90% of the emitted energy is absorbed) have been computed, together with the average percentage difference within 0.9·RCSDA and 0.9·X90 for electrons and isotopes, respectively.Results: Concerning monoenergetic electrons, within 0.8·RCSDA (where 90%–97% of the particle energy is deposed), fluka and penelope agree mostly within 7%, except for 10 and 20 keV electrons (12% in water, 8
Patni, H K; Nadar, M Y; Akar, D K; Bhati, S; Sarkar, P K
2011-11-01
The adult reference male and female computational voxel phantoms recommended by ICRP are adapted into the Monte Carlo transport code FLUKA. The FLUKA code is then utilised for computation of dose conversion coefficients (DCCs) expressed in absorbed dose per air kerma free-in-air for colon, lungs, stomach wall, breast, gonads, urinary bladder, oesophagus, liver and thyroid due to a broad parallel beam of mono-energetic photons impinging in anterior-posterior and posterior-anterior directions in the energy range of 15 keV-10 MeV. The computed DCCs of colon, lungs, stomach wall and breast are found to be in good agreement with the results published in ICRP publication 110. The present work thus validates the use of FLUKA code in computation of organ DCCs for photons using ICRP adult voxel phantoms. Further, the DCCs for gonads, urinary bladder, oesophagus, liver and thyroid are evaluated and compared with results published in ICRP 74 in the above-mentioned energy range and geometries. Significant differences in DCCs are observed for breast, testis and thyroid above 1 MeV, and for most of the organs at energies below 60 keV in comparison with the results published in ICRP 74. The DCCs of female voxel phantom were found to be higher in comparison with male phantom for almost all organs in both the geometries. PMID:21147784
Charles A. Wemple; Joshua J. Cogliati
2005-04-01
A univel geometry, neutral particle Monte Carlo transport code, written entirely in the Java programming language, is under development for medical radiotherapy applications. The code uses ENDF-VI based continuous energy cross section data in a flexible XML format. Full neutron-photon coupling, including detailed photon production and photonuclear reactions, is included. Charged particle equilibrium is assumed within the patient model so that detailed transport of electrons produced by photon interactions may be neglected. External beam and internal distributed source descriptions for mixed neutron-photon sources are allowed. Flux and dose tallies are performed on a univel basis. A four-tap, shift-register-sequence random number generator is used. Initial verification and validation testing of the basic neutron transport routines is underway. The searchlight problem was chosen as a suitable first application because of the simplicity of the physical model. Results show excellent agreement with analytic solutions. Computation times for similar numbers of histories are comparable to other neutron MC codes written in C and FORTRAN.
Comparison of Space Radiation Calculations from Deterministic and Monte Carlo Transport Codes
NASA Technical Reports Server (NTRS)
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,
Torres, Javier; Buades, Manuel J; Almansa, Julio F; Guerrero, Rafael; Lallena, Antonio M
2004-02-01
Monte Carlo calculations using the codes PENELOPE and GEANT4 have been performed to characterize the dosimetric parameters of the new 20 mm long catheter-based 32P beta source manufactured by the Guidant Corporation. The dose distribution along the transverse axis and the two-dimensional dose rate table have been calculated. Also, the dose rate at the reference point, the radial dose function, and the anisotropy function were evaluated according to the adapted TG-60 formalism for cylindrical sources. PENELOPE and GEANT4 codes were first verified against previous results corresponding to the old 27 mm Guidant 32P beta source. The dose rate at the reference point for the unsheathed 27 mm source in water was calculated to be 0.215 +/- 0.001 cGy s(-1) mCi(-1), for PENELOPE, and 0.2312 +/- 0.0008 cGy s(-1) mCi(-1), for GEANT4. For the unsheathed 20 mm source, these values were 0.2908 +/- 0.0009 cGy s(-1) mCi(-1) and 0.311 0.001 cGy s(-1) mCi(-1), respectively. Also, a comparison with the limited data available on this new source is shown. We found non-negligible differences between the results obtained with PENELOPE and GEANT4. PMID:15000615
NASA Astrophysics Data System (ADS)
Lee, Y.-K.; Brun, E.
2014-04-01
The Sodium-cooled fast neutron reactor ASTRID is currently under design and development in France. Traditional ECCO/ERANOS fast reactor code system used for ASTRID core design calculations relies on multi-group JEFF-3.1.1 data library. To gauge the use of ENDF/B-VII.0 and JEFF-3.1.1 nuclear data libraries in the fast reactor applications, two recent OECD/NEA computational benchmarks specified by Argonne National Laboratory were calculated. Using the continuous-energy TRIPOLI-4 Monte Carlo transport code, both ABR-1000 MWth MOX core and metallic (U-Pu) core were investigated. Under two different fast neutron spectra and two data libraries, ENDF/B-VII.0 and JEFF-3.1.1, reactivity impact studies were performed. Using JEFF-3.1.1 library under the BOEC (Beginning of equilibrium cycle) condition, high reactivity effects of 808 ± 17 pcm and 1208 ± 17 pcm were observed for ABR-1000 MOX core and metallic core respectively. To analyze the causes of these differences in reactivity, several TRIPOLI-4 runs using mixed data libraries feature allow us to identify the nuclides and the nuclear data accounting for the major part of the observed reactivity discrepancies.
Giantsoudi, D; Schuemann, J; Dowdell, S; Paganetti, H; Jia, X; Jiang, S
2014-06-15
Purpose: For proton radiation therapy, Monte Carlo simulation (MCS) methods are recognized as the gold-standard dose calculation approach. Although previously unrealistic due to limitations in available computing power, GPU-based applications allow MCS of proton treatment fields to be performed in routine clinical use, on time scales comparable to that of conventional pencil-beam algorithms. This study focuses on validating the results of our GPU-based code (gPMC) versus fully implemented proton therapy based MCS code (TOPAS) for clinical patient cases. Methods: Two treatment sites were selected to provide clinical cases for this study: head-and-neck cases due to anatomical geometrical complexity (air cavities and density heterogeneities), making dose calculation very challenging, and prostate cases due to higher proton energies used and close proximity of the treatment target to sensitive organs at risk. Both gPMC and TOPAS methods were used to calculate 3-dimensional dose distributions for all patients in this study. Comparisons were performed based on target coverage indices (mean dose, V90 and D90) and gamma index distributions for 2% of the prescription dose and 2mm. Results: For seven out of eight studied cases, mean target dose, V90 and D90 differed less than 2% between TOPAS and gPMC dose distributions. Gamma index analysis for all prostate patients resulted in passing rate of more than 99% of voxels in the target. Four out of five head-neck-cases showed passing rate of gamma index for the target of more than 99%, the fifth having a gamma index passing rate of 93%. Conclusion: Our current work showed excellent agreement between our GPU-based MCS code and fully implemented proton therapy based MC code for a group of dosimetrically challenging patient cases.
A Monte Carlo Code for Relativistic Radiation Transport around Kerr Black Holes
NASA Astrophysics Data System (ADS)
Schnittman, Jeremy D.; Krolik, Julian H.
2013-11-01
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.
A Monte Carlo Code for Relativistic Radiation Transport Around Kerr Black Holes
NASA Technical Reports Server (NTRS)
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.
A MONTE CARLO CODE FOR RELATIVISTIC RADIATION TRANSPORT AROUND KERR BLACK HOLES
Schnittman, Jeremy D.; Krolik, Julian H. E-mail: jhk@pha.jhu.edu
2013-11-01
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.
Liu, T.; Ding, A.; Ji, W.; Xu, X. G.; Carothers, C. D.; Brown, F. B.
2012-07-01
Monte Carlo (MC) method is able to accurately calculate eigenvalues in reactor analysis. Its lengthy computation time can be reduced by general-purpose computing on Graphics Processing Units (GPU), one of the latest parallel computing techniques under development. The method of porting a regular transport code to GPU is usually very straightforward due to the 'embarrassingly parallel' nature of MC code. However, the situation becomes different for eigenvalue calculation in that it will be performed on a generation-by-generation basis and the thread coordination should be explicitly taken care of. This paper presents our effort to develop such a GPU-based MC code in Compute Unified Device Architecture (CUDA) environment. The code is able to perform eigenvalue calculation under simple geometries on a multi-GPU system. The specifics of algorithm design, including thread organization and memory management were described in detail. The original CPU version of the code was tested on an Intel Xeon X5660 2.8 GHz CPU, and the adapted GPU version was tested on NVIDIA Tesla M2090 GPUs. Double-precision floating point format was used throughout the calculation. The result showed that a speedup of 7.0 and 33.3 were obtained for a bare spherical core and a binary slab system respectively. The speedup factor was further increased by a factor of {approx}2 on a dual GPU system. The upper limit of device-level parallelism was analyzed, and a possible method to enhance the thread-level parallelism was proposed. (authors)
Monte Carlo Predictions of Prompt Fission Neutrons and Photons: a Code Comparison
NASA Astrophysics Data System (ADS)
Talou, P.; Kawano, T.; Stetcu, I.; Vogt, R.; Randrup, J.
2014-04-01
This paper reports on initial comparisons between the LANL CGMF and LBNL/LLNL FREYA codes, which both aim at computing prompt fission neutrons and gammas. While the methodologies used in both codes are somewhat similar, the detailed implementations and physical assumptions are different. We are investigating how some of these differences impact predictions.
NASA Astrophysics Data System (ADS)
Nath, S.
2008-10-01
We present a semimicroscopic Monte Carlo code for calculating absolute transmission efficiency of recoil separators for heavy ion-induced complete fusion reactions. The code generates realistic distributions for energy, charge state and angle of evaporation residues. Residue trajectories are calculated using first order ion optical transfer matrices. Trajectory plots in the dispersive and the non-dispersive planes are generated. Using this code, we have obtained good agreement between calculated and measured transmission efficiencies for the Heavy Ion Reaction Analyzer at IUAC. The code can be adapted easily to any other electromagnetic recoil separator. Program summaryProgram title: TERS Catalogue identifier: AEBD_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEBD_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 6818 No. of bytes in distributed program, including test data, etc.: 1 216 097 Distribution format: tar.gz Programming language: C Computer: The code has been developed and tested on a PC with Intel Pentium IV processor Operating system: Linux RAM: About 8 Mbytes Classification: 17.7 External routines: pgplot graphics subroutine library [1] should be installed in the system for generating residue trajectory plots. Nature of problem: Recoil separators are employed to select and identify nuclei of interest, produced in a nuclear reaction, rejecting unreacted beam and other undesired reaction products. It is important to know what fraction of the selected nuclei, leaving the target, reaches the detection system. This information is crucial for determining absolute cross section of the studied reaction. Solution method:Interaction of projectiles with target nuclei is treated event by event, semimicroscopically. Position and angle (with respect to beam
Energy Science and Technology Software Center (ESTSC)
1991-08-01
Version: 00 The original MORSE code was a multipurpose neutron and gamma-ray transport Monte Carlo code. It was designed as a tool for solving most shielding problems. Through the use of multigroup cross sections, the solution of neutron, gamma-ray, or coupled neutron-gamma-ray problems could be obtained in either the forward or adjoint mode. Time dependence for both shielding and criticality problems is provided. General three-dimensional geometry could be used with an albedo option available atmore » any material surface. Isotropic or anisotropic scattering up to a P16 expansion of the angular distribution was allowed. MORSE-CG incorporated the Mathematical Applications, Inc. (MAGI) combinatorial geometry routines. MORSE-B modifies the Monte Carlo neutron and photon transport computer code MORSE-CG by adding routines which allow various flexible options.« less
Gurevich, M. I.; Oleynik, D. S.; Russkov, A. A.; Voloschenko, A. M.
2006-07-01
The tracing algorithm that is implemented in the geometrical module of Monte-Carlo transport code MCU is applied to calculate the volume fractions of original materials by spatial cells of the mesh that overlays problem geometry. In this way the 3D combinatorial geometry presentation of the problem geometry, used by MCU code, is transformed to the user defined 2D or 3D bit-mapped ones. Next, these data are used in the volume fraction (VF) method to approximate problem geometry by introducing additional mixtures for spatial cells, where a few original materials are included. We have found that in solving realistic 2D and 3D core problems a sufficiently fast convergence of the VF method takes place if the spatial mesh is refined. Virtually, the proposed variant of implementation of the VF method seems as a suitable geometry interface between Monte-Carlo and S{sub n} transport codes. (authors)
Mohanty, P. K.; Dugad, S. R.; Gupta, S. K.
2012-04-15
A detailed description of a compact Monte Carlo simulation code ''G3sim'' for studying the performance of a plastic scintillator detector with wavelength shifter (WLS) fiber readout is presented. G3sim was developed for optimizing the design of new scintillator detectors used in the GRAPES-3 extensive air shower experiment. Propagation of the blue photons produced by the passage of relativistic charged particles in the scintillator is treated by incorporating the absorption, total internal, and diffuse reflections. Capture of blue photons by the WLS fibers and subsequent re-emission of longer wavelength green photons is appropriately treated. The trapping and propagation of green photons inside the WLS fiber is treated using the laws of optics for meridional and skew rays. Propagation time of each photon is taken into account for the generation of the electrical signal at the photomultiplier. A comparison of the results from G3sim with the performance of a prototype scintillator detector showed an excellent agreement between the simulated and measured properties. The simulation results can be parametrized in terms of exponential functions providing a deeper insight into the functioning of these versatile detectors. G3sim can be used to aid the design and optimize the performance of scintillator detectors prior to actual fabrication that may result in a considerable saving of time, labor, and money spent.
Thiam, C; Bobin, C; Bouchard, J
2010-01-01
The implementation of the TDCR method (Triple to Double Coincidence Ratio) is based on a liquid scintillation system which comprises three photomultipliers; at LNHB, this counter can also be used in the beta-channel of a 4pi(LS)beta-gamma coincidence counting equipment. It is generally considered that the gamma-sensitivity of the liquid scintillation detector comes from the interaction of the gamma-photons in the scintillation cocktail but when introducing solid gamma-ray emitting sources instead of the scintillation vial, light emitted by the surrounding of the counter is observed. The explanation proposed in this article is that this effect comes from the emission of Cherenkov photons induced by Compton diffusion in the photomultiplier windows. In order to support this assertion, the creation and the propagation of Cherenkov photons inside the TDCR counter is simulated using the Monte Carlo code GEANT4. Stochastic calculations of double coincidences confirm the hypothesis of Cherenkov light produced in the photomultiplier windows. PMID:20031429
Mohanty, P K; Dugad, S R; Gupta, S K
2012-04-01
A detailed description of a compact Monte Carlo simulation code "G3sim" for studying the performance of a plastic scintillator detector with wavelength shifter (WLS) fiber readout is presented. G3sim was developed for optimizing the design of new scintillator detectors used in the GRAPES-3 extensive air shower experiment. Propagation of the blue photons produced by the passage of relativistic charged particles in the scintillator is treated by incorporating the absorption, total internal, and diffuse reflections. Capture of blue photons by the WLS fibers and subsequent re-emission of longer wavelength green photons is appropriately treated. The trapping and propagation of green photons inside the WLS fiber is treated using the laws of optics for meridional and skew rays. Propagation time of each photon is taken into account for the generation of the electrical signal at the photomultiplier. A comparison of the results from G3sim with the performance of a prototype scintillator detector showed an excellent agreement between the simulated and measured properties. The simulation results can be parametrized in terms of exponential functions providing a deeper insight into the functioning of these versatile detectors. G3sim can be used to aid the design and optimize the performance of scintillator detectors prior to actual fabrication that may result in a considerable saving of time, labor, and money spent. PMID:22559526
NASA Astrophysics Data System (ADS)
Ibey, Bennett L.; Lee, Seungjoon; Ericson, M. Nance; Wilson, Mark A.; Cote, Gerard L.
2004-06-01
A Multi-Layer Monte Carlo (MLMC) model was developed to predict the results of in vivo blood perfusion and oxygenation measurement of transplanted organs as measured by an indwelling optical sensor. A sensor has been developed which uses three-source excitation in the red and infrared ranges (660, 810, 940 nm). In vitro data was taken using this sensor by changing the oxygenation state of whole blood and passing it through a single-tube pump system wrapped in bovine liver tissue. The collected data showed that the red signal increased as blood oxygenation increased and infrared signal decreased. The center wavelength of 810 nanometers was shown to be quite indifferent to blood oxygenation change. A model was developed using MLMC code that sampled the wavelength range from 600-1000 nanometers every 6 nanometers. Using scattering and absorption data for blood and liver tissue within this wavelength range, a five-layer model was developed (tissue, clear tubing, blood, clear tubing, tissue). The theoretical data generated from this model was compared to the in vitro data and showed good correlation with changing blood oxygenation.
NASA Astrophysics Data System (ADS)
Kahraman, A.; Kaya, S.; Jaksic, A.; Yilmaz, E.
2015-05-01
Radiation-sensing Field Effect Transistors (RadFETs or MOSFET dosimeters) with SiO2 gate dielectric have found applications in space, radiotherapy clinics, and high-energy physics laboratories. More sensitive RadFETs, which require modifications in device design, including gate dielectric, are being considered for personal dosimetry applications. This paper presents results of a detailed study of the RadFET energy response simulated with PENELOPE Monte Carlo code. Alternative materials to SiO2 were investigated to develop high-efficiency new radiation sensors. Namely, in addition to SiO2, Al2O3 and HfO2 were simulated as gate material and deposited energy amounts in these layers were determined for photon irradiation with energies between 20 keV and 5 MeV. The simulations were performed for capped and uncapped configurations of devices irradiated by point and extended sources, the surface area of which is the same with that of the RadFETs. Energy distributions of transmitted and backscattered photons were estimated using impact detectors to provide information about particle fluxes within the geometrical structures. The absorbed energy values in the RadFETs material zones were recorded. For photons with low and medium energies, the physical processes that affect the absorbed energy values in different gate materials are discussed on the basis of modelling results. The results show that HfO2 is the most promising of the simulated gate materials.
Three-dimensional cellular dosimetry of I-131 mIBG in neuroblastoma with EGS4 Monte Carlo code
Gouriou, J.; Ricard, M.; Lumbroso, J.; Aubert, B. |
1995-05-01
The adequate distribution of radiation dose to tumor cells is the most important factor for the outcome of internal (metabolic) radiotherapy. This study investigates the dosimetry of I-131 meta-iodobenzyl-guanidine at the cellular level in neuroblastoma. We developed a program based on the EGS4 Monte Carlo code allowing the computation of basic dosimetric parameters such as absorbed and cumulated fractions, scaled dose point kernels and dose rates, especially for radionuclides with therapeutic potential. It can be applied to various types of 3-D radionuclide tumor distributions. Geometrical parameters and mIBG uptake in xenografted tumors (nude mice, SK-N-SH) were obtained from micro-autoradiographies and SIMS microscopy images. The tumor could be simulated by a spheroid (500 {mu}m in radius) made up of spherical cells (9 {mu}m in radius) with a 1 {mu}m cytoplasm. Among this cell population, only 3% bound mIBG with local maximal rates of up to 16%. The radiation doses were calculated for I-131, since this radionuclide is the most widely used for labelling mIBG for a therapeutic potential. It can be applied to various types of 3-D radionuclide tumor distributions. Geometrical parameters and mIBG uptake in xenografted tumors (nude mice, SK-N-SH) were obtained from micro-autoradiographies and SIMS microscopy images.
NASA Astrophysics Data System (ADS)
Mohanty, P. K.; Dugad, S. R.; Gupta, S. K.
2012-04-01
A detailed description of a compact Monte Carlo simulation code "G3sim" for studying the performance of a plastic scintillator detector with wavelength shifter (WLS) fiber readout is presented. G3sim was developed for optimizing the design of new scintillator detectors used in the GRAPES-3 extensive air shower experiment. Propagation of the blue photons produced by the passage of relativistic charged particles in the scintillator is treated by incorporating the absorption, total internal, and diffuse reflections. Capture of blue photons by the WLS fibers and subsequent re-emission of longer wavelength green photons is appropriately treated. The trapping and propagation of green photons inside the WLS fiber is treated using the laws of optics for meridional and skew rays. Propagation time of each photon is taken into account for the generation of the electrical signal at the photomultiplier. A comparison of the results from G3sim with the performance of a prototype scintillator detector showed an excellent agreement between the simulated and measured properties. The simulation results can be parametrized in terms of exponential functions providing a deeper insight into the functioning of these versatile detectors. G3sim can be used to aid the design and optimize the performance of scintillator detectors prior to actual fabrication that may result in a considerable saving of time, labor, and money spent.
Interpretation of 3D void measurements with Tripoli4.6/JEFF3.1.1 Monte Carlo code
Blaise, P.; Colomba, A.
2012-07-01
The present work details the first analysis of the 3D void phase conducted during the EPICURE/UM17x17/7% mixed UOX/MOX configuration. This configuration is composed of a homogeneous central 17x17 MOX-7% assembly, surrounded by portions of 17x17 1102 assemblies with guide-tubes. The void bubble is modelled by a small waterproof 5x5 fuel pin parallelepiped box of 11 cm height, placed in the centre of the MOX assembly. This bubble, initially placed at the core mid-plane, is then moved in different axial positions to study the evolution in the core of the axial perturbation. Then, to simulate the growing of this bubble in order to understand the effects of increased void fraction along the fuel pin, 3 and 5 bubbles have been stacked axially, from the core mid-plane. The C/E comparison obtained with the Monte Carlo code Tripoli4 for both radial and axial fission rate distributions, and in particular the reproduction of the very important flux gradients at the void/water interfaces, changing as the bubble is displaced along the z-axis are very satisfactory. It demonstrates both the capability of the code and its library to reproduce this kind of situation, as the very good quality of the experimental results, confirming the UM-17x17 as an excellent experimental benchmark for 3D code validation. This work has been performed within the frame of the V and V program for the future APOLL03 deterministic code of CEA starting in 2012, and its V and V benchmarking database. (authors)
NASA Astrophysics Data System (ADS)
Kim, Jung-Ha; Hill, Robin; Kuncic, Zdenka
2012-07-01
The Monte Carlo (MC) method has proven invaluable for radiation transport simulations to accurately determine radiation doses and is widely considered a reliable computational measure that can substitute a physical experiment where direct measurements are not possible or feasible. In the EGSnrc/BEAMnrc MC codes, there are several user-specified parameters and customized transport algorithms, which may affect the calculation results. In order to fully utilize the MC methods available in these codes, it is essential to understand all these options and to use them appropriately. In this study, the effects of the electron transport algorithms in EGSnrc/BEAMnrc, which are often a trade-off between calculation accuracy and efficiency, were investigated in the buildup region of a homogeneous water phantom and also in a heterogeneous phantom using the DOSRZnrc user code. The algorithms and parameters investigated include: boundary crossing algorithm (BCA), skin depth, electron step algorithm (ESA), global electron cutoff energy (ECUT) and electron production cutoff energy (AE). The variations in calculated buildup doses were found to be larger than 10% for different user-specified transport parameters. We found that using BCA = EXACT gave the best results in terms of accuracy and efficiency in calculating buildup doses using DOSRZnrc. In addition, using the ESA = PRESTA-I option was found to be the best way of reducing the total calculation time without losing accuracy in the results at high energies (few keV ∼ MeV). We also found that although choosing a higher ECUT/AE value in the beam modelling can dramatically improve computation efficiency, there is a significant trade-off in surface dose uncertainty. Our study demonstrates that a careful choice of user-specified transport parameters is required when conducting similar MC calculations.
Electron-photon transport using the EGS4 (Electron Gamma Shower) Monte Carlo Code
Nelson, W.R.; Hirayama, H.; Rogers, D.W.O.
1986-01-01
The EGS (Electron Gamma Shower) code system was formally introduced in 1978 as a package, most commonly referred to as ESG3. It was designed to simulate electromagnetic cascades in various geometries and at energies up to a few thousand gigaelectron volts and down to cutoff kinetic energies of 0.1 MeV (photons) and 1 MeV (electrons). There have been many requests to extend EGS3 down to lower energies and this is a major, but not the only, reason for creating EGS4, which is now available for general distribution and is the subject of this presentation. A summary is given of the main features of the ESG4 code system, including statements about the physics that has been put into it and what can be realistically simulated. 6 refs.
Modelling dose distribution in tubing and cable using CYLTRAN and ACCEPT Monte Carlo simulation code
Weiss, D.E.; Kensek, R.P.
1993-12-31
One of the difficulties in the irradiation of non-slab geometries, such as a tube, is the uneven penetration of the electrons. A simple model of the distribution of dose in a tube or cable in relationship to voltage, composition, wall thickness and diameter can be mapped using the cylinder geometry provided for in the ITS/CYLTRAN code, complete with automatic subzoning. The reality of more complex 3D geometry to include effects of window foil, backscattering fixtures and beam scanning angles can be more completely accounted for by using the ITS/ACCEPT code with a line source update and a system of intersecting wedges to define input zones for mapping dose distributions in a tube. Thus, all of the variables that affect dose distribution can be modelled without the need to run time consuming and costly factory experiments. The effects of composition changes on dose distribution can also be anticipated.
NASA Astrophysics Data System (ADS)
Fracchiolla, F.; Lorentini, S.; Widesott, L.; Schwarz, M.
2015-11-01
We propose a method of creating and validating a Monte Carlo (MC) model of a proton Pencil Beam Scanning (PBS) machine using only commissioning measurements and avoiding the nozzle modeling. Measurements with a scintillating screen coupled with a CCD camera, ionization chamber and a Faraday Cup were used to model the beam in TOPAS without using any machine parameter information but the virtual source distance from the isocenter. Then the model was validated on simple Spread Out Bragg Peaks (SOBP) delivered in water phantom and with six realistic clinical plans (many involving 3 or more fields) on an anthropomorphic phantom. In particular the behavior of the moveable Range Shifter (RS) feature was investigated and its modeling has been proposed. The gamma analysis (3%,3 mm) was used to compare MC, TPS (XiO-ELEKTA) and measured 2D dose distributions (using radiochromic film). The MC modeling proposed here shows good results in the validation phase, both for simple irradiation geometry (SOBP in water) and for modulated treatment fields (on anthropomorphic phantoms). In particular head lesions were investigated and both MC and TPS data were compared with measurements. Treatment plans with no RS always showed a very good agreement with both of them (γ -Passing Rate (PR) > 95%). Treatment plans in which the RS was needed were also tested and validated. For these treatment plans MC results showed better agreement with measurements (γ -PR > 93%) than the one coming from TPS (γ -PR < 88%). This work shows how to simplify the MC modeling of a PBS machine for proton therapy treatments without accounting for any hardware components and proposes a more reliable RS modeling than the one implemented in our TPS. The validation process has shown how this code is a valid candidate for a completely independent treatment plan dose calculation algorithm. This makes the code an important future tool for the patient specific QA verification process.
Gohar, Y.; Zhong, Z.; Talamo, A.; Nuclear Engineering Division
2009-06-09
Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an electron accelerator driven subcritical (ADS) facility, using the KIPT electron accelerator. The neutron source of the subcritical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and electron energy in the range of 100 to 200 MeV. The main functions of the subcritical assembly are the production of medical isotopes and the support of the Ukraine nuclear power industry. Neutron physics experiments and material structure analyses are planned using this facility. With the 100 KW electron beam power, the total thermal power of the facility is {approx}375 kW including the fission power of {approx}260 kW. The burnup of the fissile materials and the buildup of fission products reduce continuously the reactivity during the operation, which reduces the neutron flux level and consequently the facility performance. To preserve the neutron flux level during the operation, fuel assemblies should be added after long operating periods to compensate for the lost reactivity. This process requires accurate prediction of the fuel burnup, the decay behavior of the fission produces, and the introduced reactivity from adding fresh fuel assemblies. The recent developments of the Monte Carlo computer codes, the high speed capability of the computer processors, and the parallel computation techniques made it possible to perform three-dimensional detailed burnup simulations. A full detailed three-dimensional geometrical model is used for the burnup simulations with continuous energy nuclear data libraries for the transport calculations and 63-multigroup or one group cross sections libraries for the depletion calculations. Monte Carlo Computer code MCNPX and MCB are utilized for this study. MCNPX transports the
ITS Version 4.0: Electron/photon Monte Carlo transport codes
Halbleib, J.A,; Kensek, R.P.; Seltzer, S.M.
1995-07-01
The current publicly released version of the Integrated TIGER Series (ITS), Version 3.0, has been widely distributed both domestically and internationally, and feedback has been very positive. This feedback as well as our own experience have convinced us to upgrade the system in order to honor specific user requests for new features and to implement other new features that will improve the physical accuracy of the system and permit additional variance reduction. This presentation we will focus on components of the upgrade that (1) improve the physical model, (2) provide new and extended capabilities to the three-dimensional combinatorial-geometry (CG) of the ACCEPT codes, and (3) permit significant variance reduction in an important class of radiation effects applications.
Thermal neutron response of a boron-coated GEM detector via GEANT4 Monte Carlo code.
Jamil, M; Rhee, J T; Kim, H G; Ahmad, Farzana; Jeon, Y J
2014-10-22
In this work, we report the design configuration and the performance of the hybrid Gas Electron Multiplier (GEM) detector. In order to make the detector sensitive to thermal neutrons, the forward electrode of the GEM has been coated with the enriched boron-10 material, which works as a neutron converter. A total of 5×5cm(2) configuration of GEM has been used for thermal neutron studies. The response of the detector has been estimated via using GEANT4 MC code with two different physics lists. Using the QGSP_BIC_HP physics list, the neutron detection efficiency was determined to be about 3%, while with QGSP_BERT_HP physics list the efficiency was around 2.5%, at the incident thermal neutron energies of 25meV. The higher response of the detector proves that GEM-coated with boron converter improves the efficiency for thermal neutrons detection. PMID:25464183
A relativistic Monte Carlo binary collision model for use in plasma particle simulation codes
Procassini, R.J.; Birdsall, C.K.; Morse, E.C.; Cohen, B.I.
1987-05-14
Particle simulations of plasma physics phenomena employ far fewer particles than the systems which are being simulated, owing to the limited speed and memory capacity of even the most powerful supercomputers. If the simulation consists of point particles in a gridless domain, then the combination of the small number of particles in a Debye sphere and the possibility of zero-impact-parameters, large-angle scattering results in a significant enhancement of fluctuation phenomena such as collisions. Collisional processes in a simulation may be difficult because of disparate time scales. A comparison of the relevant physical time scales of the system that is being simulated usually yields a large range of values. For instance, the grid-cell transit time is usually several orders of magnitude smaller than the 90/sup 0/ scattering time. Much of the physical phenomena of interest in the simulation are due to these long-time-scale collisional processes, but short-time-scale processes (such as particle bounce times in a mirror or tokamak) must be adequately resolved if the plasma dielectric response and the plasma potential are to be accurately determined. The following paper outlines the physics and operation of the binary collision model within the electrostatic code and presents the results of computer simulations of velocity space transport which were run to test the accuracy of the model. Also discussed are the timing statistics for the collision package relative to the other major physics packages in the code, as well as recommendations on the frequency of use of the collision package within the simulation sequence.