The MC21 Monte Carlo Transport Code
International Nuclear Information System (INIS)
Sutton TM; Donovan TJ; Trumbull TH; Dobreff PS; Caro E; Griesheimer DP; Tyburski LJ; Carpenter DC; Joo H
2007-01-01
MC21 is a new Monte Carlo neutron and photon transport code currently under joint development at the Knolls Atomic Power Laboratory and the Bettis Atomic Power Laboratory. MC21 is the Monte Carlo transport kernel of the broader Common Monte Carlo Design Tool (CMCDT), which is also currently under development. The vision for CMCDT is to provide an automated, computer-aided modeling and post-processing environment integrated with a Monte Carlo solver that is optimized for reactor analysis. CMCDT represents a strategy to push the Monte Carlo method beyond its traditional role as a benchmarking tool or ''tool of last resort'' and into a dominant design role. This paper describes various aspects of the code, including the neutron physics and nuclear data treatments, the geometry representation, and the tally and depletion capabilities
Parallel processing Monte Carlo radiation transport codes
International Nuclear Information System (INIS)
McKinney, G.W.
1994-01-01
Issues related to distributed-memory multiprocessing as applied to Monte Carlo radiation transport are discussed. Measurements of communication overhead are presented for the radiation transport code MCNP which employs the communication software package PVM, and average efficiency curves are provided for a homogeneous virtual machine
Optix: A Monte Carlo scintillation light transport code
Energy Technology Data Exchange (ETDEWEB)
Safari, M.J., E-mail: mjsafari@aut.ac.ir [Department of Energy Engineering and Physics, Amir Kabir University of Technology, PO Box 15875-4413, Tehran (Iran, Islamic Republic of); Afarideh, H. [Department of Energy Engineering and Physics, Amir Kabir University of Technology, PO Box 15875-4413, Tehran (Iran, Islamic Republic of); Ghal-Eh, N. [School of Physics, Damghan University, PO Box 36716-41167, Damghan (Iran, Islamic Republic of); Davani, F. Abbasi [Nuclear Engineering Department, Shahid Beheshti University, PO Box 1983963113, Tehran (Iran, Islamic Republic of)
2014-02-11
The paper reports on the capabilities of Monte Carlo scintillation light transport code Optix, which is an extended version of previously introduced code Optics. Optix provides the user a variety of both numerical and graphical outputs with a very simple and user-friendly input structure. A benchmarking strategy has been adopted based on the comparison with experimental results, semi-analytical solutions, and other Monte Carlo simulation codes to verify various aspects of the developed code. Besides, some extensive comparisons have been made against the tracking abilities of general-purpose MCNPX and FLUKA codes. The presented benchmark results for the Optix code exhibit promising agreements. -- Highlights: • Monte Carlo simulation of scintillation light transport in 3D geometry. • Evaluation of angular distribution of detected photons. • Benchmark studies to check the accuracy of Monte Carlo simulations.
TRIPOLI-4: Monte Carlo transport code functionalities and applications
International Nuclear Information System (INIS)
Both, J.P.; Lee, Y.K.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.
2003-01-01
Tripoli-4 is a three dimensional calculations code using the Monte Carlo method to simulate the transport of neutrons, photons, electrons and positrons. This code is used in four application fields: the protection studies, the criticality studies, the core studies and the instrumentation studies. Geometry, cross sections, description of sources, principle. (N.C.)
Energy Technology Data Exchange (ETDEWEB)
Both, J P; Lee, Y K; Mazzolo, A; Peneliau, Y; Petit, O; Roesslinger, B [CEA Saclay, Dir. de l' Energie Nucleaire (DEN), Service d' Etudes de Reacteurs et de Modelisation Avancee, 91 - Gif sur Yvette (France)
2003-07-01
Tripoli-4 is a three dimensional calculations code using the Monte Carlo method to simulate the transport of neutrons, photons, electrons and positrons. This code is used in four application fields: the protection studies, the criticality studies, the core studies and the instrumentation studies. Geometry, cross sections, description of sources, principle. (N.C.)
A New Monte Carlo Neutron Transport Code at UNIST
International Nuclear Information System (INIS)
Lee, Hyunsuk; Kong, Chidong; Lee, Deokjung
2014-01-01
Monte Carlo neutron transport code named MCS is under development at UNIST for the advanced reactor design and research purpose. This MC code can be used for fixed source calculation and criticality calculation. Continuous energy neutron cross section data and multi-group cross section data can be used for the MC calculation. This paper presents the overview of developed MC code and its calculation results. The real time fixed source calculation ability is also tested in this paper. The calculation results show good agreement with commercial code and experiment. A new Monte Carlo neutron transport code is being developed at UNIST. The MC codes are tested with several benchmark problems: ICSBEP, VENUS-2, and Hoogenboom-Martin benchmark. These benchmarks covers pin geometry to 3-dimensional whole core, and results shows good agreement with reference results
Verification of Monte Carlo transport codes by activation experiments
Chetvertkova, Vera
2013-01-01
With the increasing energies and intensities of heavy-ion accelerator facilities, the problem of an excessive activation of the accelerator components caused by beam losses becomes more and more important. Numerical experiments using Monte Carlo transport codes are performed in order to assess the levels of activation. The heavy-ion versions of the codes were released approximately a decade ago, therefore the verification is needed to be sure that they give reasonable results. Present work is...
A computer code package for electron transport Monte Carlo simulation
International Nuclear Information System (INIS)
Popescu, Lucretiu M.
1999-01-01
A computer code package was developed for solving various electron transport problems by Monte Carlo simulation. It is based on condensed history Monte Carlo algorithm. In order to get reliable results over wide ranges of electron energies and target atomic numbers, specific techniques of electron transport were implemented such as: Moliere multiscatter angular distributions, Blunck-Leisegang multiscatter energy distribution, sampling of electron-electron and Bremsstrahlung individual interactions. Path-length and lateral displacement corrections algorithms and the module for computing collision, radiative and total restricted stopping powers and ranges of electrons are also included. Comparisons of simulation results with experimental measurements are finally presented. (author)
Acceleration of a Monte Carlo radiation transport code
International Nuclear Information System (INIS)
Hochstedler, R.D.; Smith, L.M.
1996-01-01
Execution time for the Integrated TIGER Series (ITS) Monte Carlo radiation transport code has been reduced by careful re-coding of computationally intensive subroutines. Three test cases for the TIGER (1-D slab geometry), CYLTRAN (2-D cylindrical geometry), and ACCEPT (3-D arbitrary geometry) codes were identified and used to benchmark and profile program execution. Based upon these results, sixteen top time-consuming subroutines were examined and nine of them modified to accelerate computations with equivalent numerical output to the original. The results obtained via this study indicate that speedup factors of 1.90 for the TIGER code, 1.67 for the CYLTRAN code, and 1.11 for the ACCEPT code are achievable. copyright 1996 American Institute of Physics
grmonty: A MONTE CARLO CODE FOR RELATIVISTIC RADIATIVE TRANSPORT
International Nuclear Information System (INIS)
Dolence, Joshua C.; Gammie, Charles F.; Leung, Po Kin; Moscibrodzka, Monika
2009-01-01
We describe a Monte Carlo radiative transport code intended for calculating spectra of hot, optically thin plasmas in full general relativity. The version we describe here is designed to model hot accretion flows in the Kerr metric and therefore incorporates synchrotron emission and absorption, and Compton scattering. The code can be readily generalized, however, to account for other radiative processes and an arbitrary spacetime. We describe a suite of test problems, and demonstrate the expected N -1/2 convergence rate, where N is the number of Monte Carlo samples. Finally, we illustrate the capabilities of the code with a model calculation, a spectrum of the slowly accreting black hole Sgr A* based on data provided by a numerical general relativistic MHD model of the accreting plasma.
Automatic modeling for the monte carlo transport TRIPOLI code
International Nuclear Information System (INIS)
Zhang Junjun; Zeng Qin; Wu Yican; Wang Guozhong; FDS Team
2010-01-01
TRIPOLI, developed by CEA, France, is Monte Carlo particle transport simulation code. It has been widely applied to nuclear physics, shielding design, evaluation of nuclear safety. However, it is time-consuming and error-prone to manually describe the TRIPOLI input file. This paper implemented bi-directional conversion between CAD model and TRIPOLI model. Its feasibility and efficiency have been demonstrated by several benchmarking examples. (authors)
The OpenMC Monte Carlo particle transport code
International Nuclear Information System (INIS)
Romano, Paul K.; Forget, Benoit
2013-01-01
Highlights: ► An open source Monte Carlo particle transport code, OpenMC, has been developed. ► Solid geometry and continuous-energy physics allow high-fidelity simulations. ► Development has focused on high performance and modern I/O techniques. ► OpenMC is capable of scaling up to hundreds of thousands of processors. ► Results on a variety of benchmark problems agree with MCNP5. -- Abstract: A new Monte Carlo code called OpenMC is currently under development at the Massachusetts Institute of Technology as a tool for simulation on high-performance computing platforms. Given that many legacy codes do not scale well on existing and future parallel computer architectures, OpenMC has been developed from scratch with a focus on high performance scalable algorithms as well as modern software design practices. The present work describes the methods used in the OpenMC code and demonstrates the performance and accuracy of the code on a variety of problems.
Verification of Monte Carlo transport codes by activation experiments
Energy Technology Data Exchange (ETDEWEB)
Chetvertkova, Vera
2012-12-18
With the increasing energies and intensities of heavy-ion accelerator facilities, the problem of an excessive activation of the accelerator components caused by beam losses becomes more and more important. Numerical experiments using Monte Carlo transport codes are performed in order to assess the levels of activation. The heavy-ion versions of the codes were released approximately a decade ago, therefore the verification is needed to be sure that they give reasonable results. Present work is focused on obtaining the experimental data on activation of the targets by heavy-ion beams. Several experiments were performed at GSI Helmholtzzentrum fuer Schwerionenforschung. The interaction of nitrogen, argon and uranium beams with aluminum targets, as well as interaction of nitrogen and argon beams with copper targets was studied. After the irradiation of the targets by different ion beams from the SIS18 synchrotron at GSI, the γ-spectroscopy analysis was done: the γ-spectra of the residual activity were measured, the radioactive nuclides were identified, their amount and depth distribution were detected. The obtained experimental results were compared with the results of the Monte Carlo simulations using FLUKA, MARS and SHIELD. The discrepancies and agreements between experiment and simulations are pointed out. The origin of discrepancies is discussed. Obtained results allow for a better verification of the Monte Carlo transport codes, and also provide information for their further development. The necessity of the activation studies for accelerator applications is discussed. The limits of applicability of the heavy-ion beam-loss criteria were studied using the FLUKA code. FLUKA-simulations were done to determine the most preferable from the radiation protection point of view materials for use in accelerator components.
Parallelization of a Monte Carlo particle transport simulation code
Hadjidoukas, P.; Bousis, C.; Emfietzoglou, D.
2010-05-01
We have developed a high performance version of the Monte Carlo particle transport simulation code MC4. The original application code, developed in Visual Basic for Applications (VBA) for Microsoft Excel, was first rewritten in the C programming language for improving code portability. Several pseudo-random number generators have been also integrated and studied. The new MC4 version was then parallelized for shared and distributed-memory multiprocessor systems using the Message Passing Interface. Two parallel pseudo-random number generator libraries (SPRNG and DCMT) have been seamlessly integrated. The performance speedup of parallel MC4 has been studied on a variety of parallel computing architectures including an Intel Xeon server with 4 dual-core processors, a Sun cluster consisting of 16 nodes of 2 dual-core AMD Opteron processors and a 200 dual-processor HP cluster. For large problem size, which is limited only by the physical memory of the multiprocessor server, the speedup results are almost linear on all systems. We have validated the parallel implementation against the serial VBA and C implementations using the same random number generator. Our experimental results on the transport and energy loss of electrons in a water medium show that the serial and parallel codes are equivalent in accuracy. The present improvements allow for studying of higher particle energies with the use of more accurate physical models, and improve statistics as more particles tracks can be simulated in low response time.
KAMCCO, a reactor physics Monte Carlo neutron transport code
International Nuclear Information System (INIS)
Arnecke, G.; Borgwaldt, H.; Brandl, V.; Lalovic, M.
1976-06-01
KAMCCO is a 3-dimensional reactor Monte Carlo code for fast neutron physics problems. Two options are available for the solution of 1) the inhomogeneous time-dependent neutron transport equation (census time scheme), and 2) the homogeneous static neutron transport equation (generation cycle scheme). The user defines the desired output, e.g. estimates of reaction rates or neutron flux integrated over specified volumes in phase space and time intervals. Such primary quantities can be arbitrarily combined, also ratios of these quantities can be estimated with their errors. The Monte Carlo techniques are mostly analogue (exceptions: Importance sampling for collision processes, ELP/MELP, Russian roulette and splitting). Estimates are obtained from the collision and track length estimators. Elastic scattering takes into account first order anisotropy in the center of mass system. Inelastic scattering is processed via the evaporation model or via the excitation of discrete levels. For the calculation of cross sections, the energy is treated as a continuous variable. They are computed by a) linear interpolation, b) from optionally Doppler broadened single level Breit-Wigner resonances or c) from probability tables (in the region of statistically distributed resonances). (orig.) [de
New features of the mercury Monte Carlo particle transport code
International Nuclear Information System (INIS)
Procassini, Richard; Brantley, Patrick; Dawson, Shawn
2010-01-01
Several new capabilities have been added to the Mercury Monte Carlo transport code over the past four years. The most important algorithmic enhancement is a general, extensible infrastructure to support source, tally and variance reduction actions. For each action, the user defines a phase space, as well as any number of responses that are applied to a specified event. Tallies are accumulated into a correlated, multi-dimensional. Cartesian-product result phase space. Our approach employs a common user interface to specify the data sets and distributions that define the phase, response and result for each action. Modifications to the particle trackers include the use of facet halos (instead of extrapolative fuzz) for robust tracking, and material interface reconstruction for use in shape overlaid meshes. Support for expected-value criticality eigenvalue calculations has also been implemented. Computer science enhancements include an in-line Python interface for user customization of problem setup and output. (author)
Development of general-purpose particle and heavy ion transport monte carlo code
International Nuclear Information System (INIS)
Iwase, Hiroshi; Nakamura, Takashi; Niita, Koji
2002-01-01
The high-energy particle transport code NMTC/JAM, which has been developed at JAERI, was improved for the high-energy heavy ion transport calculation by incorporating the JQMD code, the SPAR code and the Shen formula. The new NMTC/JAM named PHITS (Particle and Heavy-Ion Transport code System) is the first general-purpose heavy ion transport Monte Carlo code over the incident energies from several MeV/nucleon to several GeV/nucleon. (author)
Automatic modeling for the Monte Carlo transport code Geant4
International Nuclear Information System (INIS)
Nie Fanzhi; Hu Liqin; Wang Guozhong; Wang Dianxi; Wu Yican; Wang Dong; Long Pengcheng; FDS Team
2015-01-01
Geant4 is a widely used Monte Carlo transport simulation package. Its geometry models could be described in Geometry Description Markup Language (GDML), but it is time-consuming and error-prone to describe the geometry models manually. This study implemented the conversion between computer-aided design (CAD) geometry models and GDML models. This method has been Studied based on Multi-Physics Coupling Analysis Modeling Program (MCAM). The tests, including FDS-Ⅱ model, demonstrated its accuracy and feasibility. (authors)
DEFF Research Database (Denmark)
Taasti, Vicki Trier; Knudsen, Helge; Holzscheiter, Michael
2015-01-01
The Monte Carlo particle transport code SHIELD-HIT12A is designed to simulate therapeutic beams for cancer radiotherapy with fast ions. SHIELD-HIT12A allows creation of antiproton beam kernels for the treatment planning system TRiP98, but first it must be benchmarked against experimental data. An...
International Nuclear Information System (INIS)
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
Study on MPI/OpenMP hybrid parallelism for Monte Carlo neutron transport code
International Nuclear Information System (INIS)
Liang Jingang; Xu Qi; Wang Kan; Liu Shiwen
2013-01-01
Parallel programming with mixed mode of messages-passing and shared-memory has several advantages when used in Monte Carlo neutron transport code, such as fitting hardware of distributed-shared clusters, economizing memory demand of Monte Carlo transport, improving parallel performance, and so on. MPI/OpenMP hybrid parallelism was implemented based on a one dimension Monte Carlo neutron transport code. Some critical factors affecting the parallel performance were analyzed and solutions were proposed for several problems such as contention access, lock contention and false sharing. After optimization the code was tested finally. It is shown that the hybrid parallel code can reach good performance just as pure MPI parallel program, while it saves a lot of memory usage at the same time. Therefore hybrid parallel is efficient for achieving large-scale parallel of Monte Carlo neutron transport. (authors)
A 3D Monte Carlo code for plasma transport in island divertors
International Nuclear Information System (INIS)
Feng, Y.; Sardei, F.; Kisslinger, J.; Grigull, P.
1997-01-01
A fully 3D self-consistent Monte Carlo code EMC3 (edge Monte Carlo 3D) for modelling the plasma transport in island divertors has been developed. In a first step, the code solves a simplified version of the 3D time-independent plasma fluid equations. Coupled to the neutral transport code EIRENE, the EMC3 code has been used to study the particle, energy and neutral transport in W7-AS island divertor configurations. First results are compared with data from different diagnostics (Langmuir probes, H α cameras and thermography). (orig.)
TOPIC: a debugging code for torus geometry input data of Monte Carlo transport code
International Nuclear Information System (INIS)
Iida, Hiromasa; Kawasaki, Hiromitsu.
1979-06-01
TOPIC has been developed for debugging geometry input data of the Monte Carlo transport code. the code has the following features: (1) It debugs the geometry input data of not only MORSE-GG but also MORSE-I capable of treating torus geometry. (2) Its calculation results are shown in figures drawn by Plotter or COM, and the regions not defined or doubly defined are easily detected. (3) It finds a multitude of input data errors in a single run. (4) The input data required in this code are few, so that it is readily usable in a time sharing system of FACOM 230-60/75 computer. Example TOPIC calculations in design study of tokamak fusion reactors (JXFR, INTOR-J) are presented. (author)
Memory bottlenecks and memory contention in multi-core Monte Carlo transport codes
International Nuclear Information System (INIS)
Tramm, J.R.; Siegel, A.R.
2013-01-01
The simulation of whole nuclear cores through the use of Monte Carlo codes requires an impracticably long time-to-solution. We have extracted a kernel that executes only the most computationally expensive steps of the Monte Carlo particle transport algorithm - the calculation of macroscopic cross sections - in an effort to expose bottlenecks within multi-core, shared memory architectures. (authors)
General-purpose Monte Carlo codes for neutron and photon transport calculations. MVP version 3
International Nuclear Information System (INIS)
Nagaya, Yasunobu
2017-01-01
JAEA has developed a general-purpose neutron/photon transport Monte Carlo code MVP. This paper describes the recent development of the MVP code and reviews the basic features and capabilities. In addition, capabilities implemented in Version 3 are also described. (author)
A vectorized Monte Carlo code for modeling photon transport in SPECT
International Nuclear Information System (INIS)
Smith, M.F.; Floyd, C.E. Jr.; Jaszczak, R.J.
1993-01-01
A vectorized Monte Carlo computer code has been developed for modeling photon transport in single photon emission computed tomography (SPECT). The code models photon transport in a uniform attenuating region and photon detection by a gamma camera. It is adapted from a history-based Monte Carlo code in which photon history data are stored in scalar variables and photon histories are computed sequentially. The vectorized code is written in FORTRAN77 and uses an event-based algorithm in which photon history data are stored in arrays and photon history computations are performed within DO loops. The indices of the DO loops range over the number of photon histories, and these loops may take advantage of the vector processing unit of our Stellar GS1000 computer for pipelined computations. Without the use of the vector processor the event-based code is faster than the history-based code because of numerical optimization performed during conversion to the event-based algorithm. When only the detection of unscattered photons is modeled, the event-based code executes 5.1 times faster with the use of the vector processor than without; when the detection of scattered and unscattered photons is modeled the speed increase is a factor of 2.9. Vectorization is a valuable way to increase the performance of Monte Carlo code for modeling photon transport in SPECT
Present status of transport code development based on Monte Carlo method
International Nuclear Information System (INIS)
Nakagawa, Masayuki
1985-01-01
The present status of development in Monte Carlo code is briefly reviewed. The main items are the followings; Application fields, Methods used in Monte Carlo code (geometry spectification, nuclear data, estimator and variance reduction technique) and unfinished works, Typical Monte Carlo codes and Merits of continuous energy Monte Carlo code. (author)
International Nuclear Information System (INIS)
Zazula, J.M.
1983-01-01
The general purpose code BALTORO was written for coupling the three-dimensional Monte-Carlo /MC/ with the one-dimensional Discrete Ordinates /DO/ radiation transport calculations. The quantity of a radiation-induced /neutrons or gamma-rays/ nuclear effect or the score from a radiation-yielding nuclear effect can be analysed in this way. (author)
ITS - The integrated TIGER series of coupled electron/photon Monte Carlo transport codes
International Nuclear Information System (INIS)
Halbleib, J.A.; Mehlhorn, T.A.
1985-01-01
The TIGER series of time-independent coupled electron/photon Monte Carlo transport codes is a group of multimaterial, multidimensional codes designed to provide a state-of-the-art description of the production and transport of the electron/photon cascade. The codes follow both electrons and photons from 1.0 GeV down to 1.0 keV, and the user has the option of combining the collisional transport with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence. Source particles can be either electrons or photons. The most important output data are (a) charge and energy deposition profiles, (b) integral and differential escape coefficients for both electrons and photons, (c) differential electron and photon flux, and (d) pulse-height distributions for selected regions of the problem geometry. The base codes of the series differ from one another primarily in their dimensionality and geometric modeling. They include (a) a one-dimensional multilayer code, (b) a code that describes the transport in two-dimensional axisymmetric cylindrical material geometries with a fully three-dimensional description of particle trajectories, and (c) a general three-dimensional transport code which employs a combinatorial geometry scheme. These base codes were designed primarily for describing radiation transport for those situations in which the detailed atomic structure of the transport medium is not important. For some applications, it is desirable to have a more detailed model of the low energy transport. The system includes three additional codes that contain a more elaborate ionization/relaxation model than the base codes. Finally, the system includes two codes that combine the collisional transport of the multidimensional base codes with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence
Progress on RMC: a Monte Carlo neutron transport code for reactor analysis
International Nuclear Information System (INIS)
Wang, Kan; Li, Zeguang; She, Ding; Liu, Yuxuan; Xu, Qi; Shen, Huayun; Yu, Ganglin
2011-01-01
This paper presents a new 3-D Monte Carlo neutron transport code named RMC (Reactor Monte Carlo code), specifically intended for reactor physics analysis. This code is being developed by Department of Engineering Physics in Tsinghua University and written in C++ and Fortran 90 language with the latest version of RMC 2.5.0. The RMC code uses the method known as the delta-tracking method to simulate neutron transport, the advantages of which include fast simulation in complex geometries and relatively simple handling of complicated geometrical objects. Some other techniques such as computational-expense oriented method and hash-table method have been developed and implemented in RMC to speedup the calculation. To meet the requirements of reactor analysis, the RMC code has the calculational functions including criticality calculation, burnup calculation and also kinetics simulation. In this paper, comparison calculations of criticality problems, burnup problems and transient problems are carried out using RMC code and other Monte Carlo codes, and the results show that RMC performs quite well in these kinds of problems. Based on MPI, RMC succeeds in parallel computation and represents a high speed-up. This code is still under intensive development and the further work directions are mentioned at the end of this paper. (author)
Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)
International Nuclear Information System (INIS)
Kirk, B.L.; West, J.T.
1984-06-01
The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided
Françoise Benz
2006-01-01
2005-2006 ACADEMIC TRAINING PROGRAMME LECTURE SERIES 27, 28, 29 June 11:00-12:00 - TH Conference Room, bldg. 4 The use of Monte Carlo radiation transport codes in radiation physics and dosimetry F. Salvat Gavalda,Univ. de Barcelona, A. FERRARI, CERN-AB, M. SILARI, CERN-SC Lecture 1. Transport and interaction of electromagnetic radiation F. Salvat Gavalda,Univ. de Barcelona Interaction models and simulation schemes implemented in modern Monte Carlo codes for the simulation of coupled electron-photon transport will be briefly reviewed. Different schemes for simulating electron transport will be discussed. Condensed algorithms, which rely on multiple-scattering theories, are comparatively fast, but less accurate than mixed algorithms, in which hard interactions (with energy loss or angular deflection larger than certain cut-off values) are simulated individually. The reliability, and limitations, of electron-interaction models and multiple-scattering theories will be analyzed. Benchmark comparisons of simu...
Design of tallying function for general purpose Monte Carlo particle transport code JMCT
International Nuclear Information System (INIS)
Shangguan Danhua; Li Gang; Deng Li; Zhang Baoyin
2013-01-01
A new postponed accumulation algorithm was proposed. Based on JCOGIN (J combinatorial geometry Monte Carlo transport infrastructure) framework and the postponed accumulation algorithm, the tallying function of the general purpose Monte Carlo neutron-photon transport code JMCT was improved markedly. JMCT gets a higher tallying efficiency than MCNP 4C by 28% for simple geometry model, and JMCT is faster than MCNP 4C by two orders of magnitude for complicated repeated structure model. The available ability of tallying function for JMCT makes firm foundation for reactor analysis and multi-step burnup calculation. (authors)
Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes
International Nuclear Information System (INIS)
Smith, L.M.; Hochstedler, R.D.
1997-01-01
Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code)
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).
Parallel processing of Monte Carlo code MCNP for particle transport problem
Energy Technology Data Exchange (ETDEWEB)
Higuchi, Kenji; Kawasaki, Takuji
1996-06-01
It is possible to vectorize or parallelize Monte Carlo codes (MC code) for photon and neutron transport problem, making use of independency of the calculation for each particle. Applicability of existing MC code to parallel processing is mentioned. As for parallel computer, we have used both vector-parallel processor and scalar-parallel processor in performance evaluation. We have made (i) vector-parallel processing of MCNP code on Monte Carlo machine Monte-4 with four vector processors, (ii) parallel processing on Paragon XP/S with 256 processors. In this report we describe the methodology and results for parallel processing on two types of parallel or distributed memory computers. In addition, we mention the evaluation of parallel programming environments for parallel computers used in the present work as a part of the work developing STA (Seamless Thinking Aid) Basic Software. (author)
PyMercury: Interactive Python for the Mercury Monte Carlo Particle Transport Code
International Nuclear Information System (INIS)
Iandola, F.N.; O'Brien, M.J.; Procassini, R.J.
2010-01-01
Monte Carlo particle transport applications are often written in low-level languages (C/C++) for optimal performance on clusters and supercomputers. However, this development approach often sacrifices straightforward usability and testing in the interest of fast application performance. To improve usability, some high-performance computing applications employ mixed-language programming with high-level and low-level languages. In this study, we consider the benefits of incorporating an interactive Python interface into a Monte Carlo application. With PyMercury, a new Python extension to the Mercury general-purpose Monte Carlo particle transport code, we improve application usability without diminishing performance. In two case studies, we illustrate how PyMercury improves usability and simplifies testing and validation in a Monte Carlo application. In short, PyMercury demonstrates the value of interactive Python for Monte Carlo particle transport applications. In the future, we expect interactive Python to play an increasingly significant role in Monte Carlo usage and testing.
Design of sampling tools for Monte Carlo particle transport code JMCT
International Nuclear Information System (INIS)
Shangguan Danhua; Li Gang; Zhang Baoyin; Deng Li
2012-01-01
A class of sampling tools for general Monte Carlo particle transport code JMCT is designed. Two ways are provided to sample from distributions. One is the utilization of special sampling methods for special distribution; the other is the utilization of general sampling methods for arbitrary discrete distribution and one-dimensional continuous distribution on a finite interval. Some open source codes are included in the general sampling method for the maximum convenience of users. The sampling results show sampling correctly from distribution which are popular in particle transport can be achieved with these tools, and the user's convenience can be assured. (authors)
Premar-2: a Monte Carlo code for radiative transport simulation in atmospheric environments
International Nuclear Information System (INIS)
Cupini, E.
1999-01-01
The peculiarities of the PREMAR-2 code, aimed at radiation transport Monte Carlo simulation in atmospheric environments in the infrared-ultraviolet frequency range, are described. With respect to the previously developed PREMAR code, besides plane multilayers, spherical multilayers and finite sequences of vertical layers, each one with its own atmospheric behaviour, are foreseen in the new code, together with the refraction phenomenon, so that long range, highly slanted paths can now be more faithfully taken into account. A zenithal angular dependence of the albedo coefficient has moreover been introduced. Lidar systems, with spatially independent source and telescope, are allowed again to be simulated, and, in this latest version of the code, sensitivity analyses to be performed. According to this last feasibility, consequences on radiation transport of small perturbations in physical components of the atmospheric environment may be analyze and the related effects on searched results estimated. The availability of a library of physical data (reaction coefficients, phase functions and refraction indexes) is required by the code, providing the essential features of the environment of interest needed of the Monte Carlo simulation. Variance reducing techniques have been enhanced in the Premar-2 code, by introducing, for instance, a local forced collision technique, especially apt to be used in Lidar system simulations. Encouraging comparisons between code and experimental results carried out at the Brasimone Centre of ENEA, have so far been obtained, even if further checks of the code are to be performed [it
Srna - Monte Carlo codes for proton transport simulation in combined and voxelized geometries
Directory of Open Access Journals (Sweden)
Ilić Radovan D.
2002-01-01
Full Text Available This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtained through the PETRA and GEANT programs. The simulation of the proton beam characterization by means of the Multi-Layer Faraday Cup and spatial distribution of positron emitters obtained by our program indicate the imminent application of Monte Carlo techniques in clinical practice.
The use of Monte Carlo radiation transport codes in radiation physics and dosimetry
CERN. Geneva; Ferrari, Alfredo; Silari, Marco
2006-01-01
Transport and interaction of electromagnetic radiation Interaction models and simulation schemes implemented in modern Monte Carlo codes for the simulation of coupled electron-photon transport will be briefly reviewed. In these codes, photon transport is simulated by using the detailed scheme, i.e., interaction by interaction. Detailed simulation is easy to implement, and the reliability of the results is only limited by the accuracy of the adopted cross sections. Simulations of electron and positron transport are more difficult, because these particles undergo a large number of interactions in the course of their slowing down. Different schemes for simulating electron transport will be discussed. Condensed algorithms, which rely on multiple-scattering theories, are comparatively fast, but less accurate than mixed algorithms, in which hard interactions (with energy loss or angular deflection larger than certain cut-off values) are simulated individually. The reliability, and limitations, of electron-interacti...
MC++: A parallel, portable, Monte Carlo neutron transport code in C++
International Nuclear Information System (INIS)
Lee, S.R.; Cummings, J.C.; Nolen, S.D.
1997-01-01
MC++ is an implicit multi-group Monte Carlo neutron transport code written in C++ and based on the Parallel Object-Oriented Methods and Applications (POOMA) class library. MC++ runs in parallel on and is portable to a wide variety of platforms, including MPPs, SMPs, and clusters of UNIX workstations. MC++ is being developed to provide transport capabilities to the Accelerated Strategic Computing Initiative (ASCI). It is also intended to form the basis of the first transport physics framework (TPF), which is a C++ class library containing appropriate abstractions, objects, and methods for the particle transport problem. The transport problem is briefly described, as well as the current status and algorithms in MC++ for solving the transport equation. The alpha version of the POOMA class library is also discussed, along with the implementation of the transport solution algorithms using POOMA. Finally, a simple test problem is defined and performance and physics results from this problem are discussed on a variety of platforms
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.
A portable, parallel, object-oriented Monte Carlo neutron transport code in C++
International Nuclear Information System (INIS)
Lee, S.R.; Cummings, J.C.; Nolen, S.D.
1997-01-01
We have developed a multi-group Monte Carlo neutron transport code using C++ and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and α-eigenvalues and is portable to and runs parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities of MC++ are discussed, along with physics and performance results on a variety of hardware, including all Accelerated Strategic Computing Initiative (ASCI) hardware. Current parallel performance indicates the ability to compute α-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed
International Nuclear Information System (INIS)
Kotegawa, Hiroshi; Sasamoto, Nobuo; Tanaka, Shun-ichi
1987-02-01
Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)
Solution of charged particle transport equation by Monte-Carlo method in the BRANDZ code system
International Nuclear Information System (INIS)
Artamonov, S.N.; Androsenko, P.A.; Androsenko, A.A.
1992-01-01
Consideration is given to the issues of Monte-Carlo employment for the solution of charged particle transport equation and its implementation in the BRANDZ code system under the conditions of real 3D geometry and all the data available on radiation-to-matter interaction in multicomponent and multilayer targets. For the solution of implantation problem the results of BRANDZ data comparison with the experiments and calculations by other codes in complexes systems are presented. The results of direct nuclear pumping process simulation for laser-active media by a proton beam are also included. 4 refs.; 7 figs
Chapoutier, Nicolas; Mollier, François; Nolin, Guillaume; Culioli, Matthieu; Mace, Jean-Reynald
2017-09-01
In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics). Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition) has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.
Directory of Open Access Journals (Sweden)
Chapoutier Nicolas
2017-01-01
Full Text Available In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics. Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.
Tripoli-4, a three-dimensional poly-kinetic particle transport Monte-Carlo code
International Nuclear Information System (INIS)
Both, J.P.; Lee, Y.K.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.; Soldevila, M.
2003-01-01
In this updated of the Monte-Carlo transport code Tripoli-4, we list and describe its current main features. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. Tripoli-4 enables the user to compute the following physical quantities: a flux, a multiplication factor, a current, a reaction rate, a dose equivalent rate as well as deposit of energy and recoil energies. For each interesting physical quantity, a Monte-Carlo simulation offers different types of estimators. Tripoli-4 has support for execution in parallel mode. Special features and applications are also presented
Tripoli-4, a three-dimensional poly-kinetic particle transport Monte-Carlo code
Energy Technology Data Exchange (ETDEWEB)
Both, J P; Lee, Y K; Mazzolo, A; Peneliau, Y; Petit, O; Roesslinger, B; Soldevila, M [CEA Saclay, Dir. de l' Energie Nucleaire (DEN/DM2S/SERMA/LEPP), 91 - Gif sur Yvette (France)
2003-07-01
In this updated of the Monte-Carlo transport code Tripoli-4, we list and describe its current main features. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. Tripoli-4 enables the user to compute the following physical quantities: a flux, a multiplication factor, a current, a reaction rate, a dose equivalent rate as well as deposit of energy and recoil energies. For each interesting physical quantity, a Monte-Carlo simulation offers different types of estimators. Tripoli-4 has support for execution in parallel mode. Special features and applications are also presented.
International Nuclear Information System (INIS)
Pandya, Tara M.; Johnson, Seth R.; Evans, Thomas M.; Davidson, Gregory G.; Hamilton, Steven P.; Godfrey, Andrew T.
2015-01-01
This paper discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package developed and maintained at Oak Ridge National Laboratory. It has been developed to scale well from laptop to small computing clusters to advanced supercomputers. Special features of Shift include hybrid capabilities for variance reduction such as CADIS and FW-CADIS, and advanced parallel decomposition and tally methods optimized for scalability on supercomputing architectures. Shift has been validated and verified against various reactor physics benchmarks and compares well to other state-of-the-art Monte Carlo radiation transport codes such as MCNP5, CE KENO-VI, and OpenMC. Some specific benchmarks used for verification and validation include the CASL VERA criticality test suite and several Westinghouse AP1000 ® problems. These benchmark and scaling studies show promising results
Srna-Monte Carlo codes for proton transport simulation in combined and voxelized geometries
Ilic, R D; Stankovic, S J
2002-01-01
This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D) dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtaine...
ITS Version 6 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes.
Energy Technology Data Exchange (ETDEWEB)
Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William
2008-04-01
ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of lineartime-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 90. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.
International Nuclear Information System (INIS)
Deng Li; Xie Zhongsheng
1999-01-01
The coupled neutron and photon transport Monte Carlo code MCNP (version 3B) has been parallelized in parallel virtual machine (PVM) and message passing interface (MPI) by modifying a previous serial code. The new code has been verified by solving sample problems. The speedup increases linearly with the number of processors and the average efficiency is up to 99% for 12-processor. (author)
PBMC: Pre-conditioned Backward Monte Carlo code for radiative transport in planetary atmospheres
García Muñoz, A.; Mills, F. P.
2017-08-01
PBMC (Pre-Conditioned Backward Monte Carlo) solves the vector Radiative Transport Equation (vRTE) and can be applied to planetary atmospheres irradiated from above. The code builds the solution by simulating the photon trajectories from the detector towards the radiation source, i.e. in the reverse order of the actual photon displacements. In accounting for the polarization in the sampling of photon propagation directions and pre-conditioning the scattering matrix with information from the scattering matrices of prior (in the BMC integration order) photon collisions, PBMC avoids the unstable and biased solutions of classical BMC algorithms for conservative, optically-thick, strongly-polarizing media such as Rayleigh atmospheres.
A user's manual for the three-dimensional Monte Carlo transport code SPARTAN
International Nuclear Information System (INIS)
Bending, R.C.; Heffer, P.J.H.
1975-09-01
SPARTAN is a general-purpose Monte Carlo particle transport code intended for neutron or gamma transport problems in reactor physics, health physics, shielding, and safety studies. The code used a very general geometry system enabling a complex layout to be described and allows the user to obtain physics data from a number of different types of source library. Special tracking and scoring techniques are used to improve the quality of the results obtained. To enable users to run SPARTAN, brief descriptions of the facilities available in the code are given and full details of data input and job control language, as well as examples of complete calculations, are included. It is anticipated that changes may be made to SPARTAN from time to time, particularly in those parts of the code which deal with physics data processing. The load module is identified by a version number and implementation date, and updates of sections of this manual will be issued when significant changes are made to the code. (author)
Penelope-2006: a code system for Monte Carlo simulation of electron and photon transport
International Nuclear Information System (INIS)
2006-01-01
The computer code system PENELOPE (version 2006) performs Monte Carlo simulation of coupled electron-photon transport in arbitrary materials for a wide energy range, from a few hundred eV to about 1 GeV. Photon transport is simulated by means of the standard, detailed simulation scheme. Electron and positron histories are generated on the basis of a mixed procedure, which combines detailed simulation of hard events with condensed simulation of soft interactions. A geometry package called PENGEOM permits the generation of random electron-photon showers in material systems consisting of homogeneous bodies limited by quadric surfaces, i.e. planes, spheres, cylinders, etc. This report is intended not only to serve as a manual of the PENELOPE code system, but also to provide the user with the necessary information to understand the details of the Monte Carlo algorithm. These proceedings contain the corresponding manual and teaching notes of the PENELOPE-2006 workshop and training course, held on 4-7 July 2006 in Barcelona, Spain. (author)
Load balancing in highly parallel processing of Monte Carlo code for particle transport
International Nuclear Information System (INIS)
Higuchi, Kenji; Takemiya, Hiroshi; Kawasaki, Takuji
1998-01-01
In parallel processing of Monte Carlo (MC) codes for neutron, photon and electron transport problems, particle histories are assigned to processors making use of independency of the calculation for each particle. Although we can easily parallelize main part of a MC code by this method, it is necessary and practically difficult to optimize the code concerning load balancing in order to attain high speedup ratio in highly parallel processing. In fact, the speedup ratio in the case of 128 processors remains in nearly one hundred times when using the test bed for the performance evaluation. Through the parallel processing of the MCNP code, which is widely used in the nuclear field, it is shown that it is difficult to attain high performance by static load balancing in especially neutron transport problems, and a load balancing method, which dynamically changes the number of assigned particles minimizing the sum of the computational and communication costs, overcomes the difficulty, resulting in nearly fifteen percentage of reduction for execution time. (author)
International Nuclear Information System (INIS)
Popescu, Lucretiu M.
2000-01-01
A computer code package (PTSIM) for particle transport Monte Carlo simulation was developed using object oriented techniques of design and programming. A flexible system for simulation of coupled photon, electron transport, facilitating development of efficient simulation applications, was obtained. For photons: Compton and photo-electric effects, pair production and Rayleigh interactions are simulated, while for electrons, a class II condensed history scheme was considered, in which catastrophic interactions (Moeller electron-electron interaction, bremsstrahlung, etc.) are treated in detail and all other interactions with reduced individual effect on electron history are grouped together using continuous slowing down approximation and energy straggling theories. Electron angular straggling is simulated using Moliere theory or a mixed model in which scatters at large angles are treated as distinct events. Comparisons with experimentally benchmarks for electron transmission and bremsstrahlung emissions energy and angular spectra, and for dose calculations are presented
International Nuclear Information System (INIS)
Homma, Y.; Moriwaki, H.; Ikeda, K.; Ohdi, S.
2013-01-01
This paper deals with the verification of the 3 dimensional triangular prismatic discrete ordinates transport calculation code ENSEMBLE-TRIZ by comparison with the 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 the beginning of cycle of an initial core and at the beginning and the end of cycle of an equilibrium core. According to the calculations, the differences between the two methodologies are smaller than 0.0002 Δk in the multiplication factor, relatively about 1% in the control rod reactivity, and 1% in the sodium void reactivity. (authors)
MCNP: a general Monte Carlo code for neutron and photon transport
International Nuclear Information System (INIS)
1979-11-01
The general-purpose Monte Carlo code MCNP ca be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation are accounted for. Thermal neutrons are described by both the free-gas and S(α,β) models. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. MCNP includes an elaborate, interactive plotting capability that allows the user to view his input geometry to help check for setup errors. Standard features which are available to improve computational efficiency include geometry splitting and Russian roulette, weight cutoff with Russian roulette, correlated sampling, analog capture or capture by weight reduction, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point or ring detectors, deterministically transporting pseudo-particles to designated regions, track-length estimators, source biasing, and several parameter cutoffs. Extensive summary information is provided to help the user better understand the physics and Monte Carlo simulation of his problem. The standard, user-defined output of MCNP includes two-way current as a function of direction across any set of surfaces or surface segments in the problem. Flux across any set of surfaces or surface segments is available. 58 figures, 28 tables
International Nuclear Information System (INIS)
Chauvet, Y.
1985-01-01
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 neutron transport problems, the authors briefly describe the work done in order to get a vector code. Vectorization principles are 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 are 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 they propose a method that is very X-MP oriented in order to get the best speedup factor on such a computer. In conclusion they prove that Monte Carlo algorithms are very well suited to future vector and parallel computers
International Nuclear Information System (INIS)
Chauvet, Y.
1985-01-01
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. (orig.)
Space applications of the MITS electron-photon Monte Carlo transport code system
International Nuclear Information System (INIS)
Kensek, R.P.; Lorence, L.J.; Halbleib, J.A.; Morel, J.E.
1996-01-01
The MITS multigroup/continuous-energy electron-photon Monte Carlo transport code system has matured to the point that it is capable of addressing more realistic three-dimensional adjoint applications. It is first employed to efficiently predict point doses as a function of source energy for simple three-dimensional experimental geometries exposed to simulated uniform isotropic planar sources of monoenergetic electrons up to 4.0 MeV. Results are in very good agreement with experimental data. It is then used to efficiently simulate dose to a detector in a subsystem of a GPS satellite due to its natural electron environment, employing a relatively complex model of the satellite. The capability for survivability analysis of space systems is demonstrated, and results are obtained with and without variance reduction
MCPT: A Monte Carlo code for simulation of photon transport in tomographic scanners
International Nuclear Information System (INIS)
Prettyman, T.H.; Gardner, R.P.; Verghese, K.
1990-01-01
MCPT is a special-purpose Monte Carlo code designed to simulate photon transport in tomographic scanners. Variance reduction schemes and sampling games present in MCPT were selected to characterize features common to most tomographic scanners. Combined splitting and biasing (CSB) games are used to systematically sample important detection pathways. An efficient splitting game is used to tally particle energy deposition in detection zones. The pulse height distribution of each detector can be found by convolving the calculated energy deposition distribution with the detector's resolution function. A general geometric modelling package, HERMETOR, is used to describe the geometry of the tomographic scanners and provide MCPT information needed for particle tracking. MCPT's modelling capabilites are described and preliminary experimental validation is presented. (orig.)
Tripoli-3: monte Carlo transport code for neutral particles - version 3.5 - users manual
International Nuclear Information System (INIS)
Vergnaud, Th.; Nimal, J.C.; Chiron, M.
2001-01-01
The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)
A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom
International Nuclear Information System (INIS)
Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H.
2014-08-01
As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)
A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom
Energy Technology Data Exchange (ETDEWEB)
Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H., E-mail: mbellezzo@gmail.br [Instituto de Pesquisas Energeticas e Nucleares / CNEN, Av. Lineu Prestes 2242, Cidade Universitaria, 05508-000 Sao Paulo (Brazil)
2014-08-15
As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)
International Nuclear Information System (INIS)
Cullen, D.E
2000-01-01
TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files
Cullen, D
2000-01-01
TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files.
International Nuclear Information System (INIS)
Choi, Sung Hoon; Kwark, Min Su; Shim, Hyung Jin
2012-01-01
As The Monte Carlo (MC) particle transport analysis for a complex system such as research reactor, accelerator, and fusion facility may require accurate modeling of the complicated geometry. Its manual modeling by using the text interface of a MC code to define the geometrical objects is tedious, lengthy and error-prone. This problem can be overcome by taking advantage of modeling capability of the computer aided design (CAD) system. There have been two kinds of approaches to develop MC code systems utilizing the CAD data: the external format conversion and the CAD kernel imbedded MC simulation. The first approach includes several interfacing programs such as McCAD, MCAM, GEOMIT etc. which were developed to automatically convert the CAD data into the MCNP geometry input data. This approach makes the most of the existing MC codes without any modifications, but implies latent data inconsistency due to the difference of the geometry modeling system. In the second approach, a MC code utilizes the CAD data for the direct particle tracking or the conversion to an internal data structure of the constructive solid geometry (CSG) and/or boundary representation (B-rep) modeling with help of a CAD kernel. MCNP-BRL and OiNC have demonstrated their capabilities of the CAD-based MC simulations. Recently we have developed a CAD-based geometry processing module for the MC particle simulation by using the OpenCASCADE (OCC) library. In the developed module, CAD data can be used for the particle tracking through primitive CAD surfaces (hereafter the CAD-based tracking) or the internal conversion to the CSG data structure. In this paper, the performances of the text-based model, the CAD-based tracking, and the internal CSG conversion are compared by using an in-house MC code, McSIM, equipped with the developed CAD-based geometry processing module
Aurora T: a Monte Carlo code for transportation of neutral atoms in a toroidal plasma
International Nuclear Information System (INIS)
Bignami, A.; Chiorrini, R.
1982-01-01
This paper contains a short description of Aurora code. This code have been developed at Princeton with Monte Carlo method for calculating neutral gas in cylindrical plasma. In this work subroutines such one can take in account toroidal geometry are developed
MCNP: a general Monte Carlo code for neutron and photon transport. Version 3A. Revision 2
International Nuclear Information System (INIS)
Briesmeister, J.F.
1986-09-01
This manual is a practical guide for the use of our general-purpose Monte Carlo code MCNP. The first chapter is a primer for the novice user. The second chapter describes the mathematics, data, physics, and Monte Carlo simulation found in MCNP. This discussion is not meant to be exhaustive - details of the particular techniques and of the Monte Carlo method itself will have to be found elsewhere. The third chapter shows the user how to prepare input for the code. The fourth chapter contains several examples, and the fifth chapter explains the output. The appendices show how to use MCNP on particular computer systems at the Los Alamos National Laboratory and also give details about some of the code internals that those who wish to modify the code may find useful. 57 refs
Premar-2: a Monte Carlo code for radiative transport simulation in atmospheric environments
Energy Technology Data Exchange (ETDEWEB)
Cupini, E. [ENEA, Centro Ricerche Ezio Clementel, Bologna, (Italy). Dipt. Innovazione
1999-07-01
The peculiarities of the PREMAR-2 code, aimed at radiation transport Monte Carlo simulation in atmospheric environments in the infrared-ultraviolet frequency range, are described. With respect to the previously developed PREMAR code, besides plane multilayers, spherical multilayers and finite sequences of vertical layers, each one with its own atmospheric behaviour, are foreseen in the new code, together with the refraction phenomenon, so that long range, highly slanted paths can now be more faithfully taken into account. A zenithal angular dependence of the albedo coefficient has moreover been introduced. Lidar systems, with spatially independent source and telescope, are allowed again to be simulated, and, in this latest version of the code, sensitivity analyses to be performed. According to this last feasibility, consequences on radiation transport of small perturbations in physical components of the atmospheric environment may be analyze and the related effects on searched results estimated. The availability of a library of physical data (reaction coefficients, phase functions and refraction indexes) is required by the code, providing the essential features of the environment of interest needed of the Monte Carlo simulation. Variance reducing techniques have been enhanced in the Premar-2 code, by introducing, for instance, a local forced collision technique, especially apt to be used in Lidar system simulations. Encouraging comparisons between code and experimental results carried out at the Brasimone Centre of ENEA, have so far been obtained, even if further checks of the code are to be performed. [Italian] Nel presente rapporto vengono descritte le principali caratteristiche del codice di calcolo PREMAR-2, che esegue la simulazione Montecarlo del trasporto della radiazione elettromagnetica nell'atmosfera, nell'intervallo di frequenza che va dall'infrarosso all'ultravioletto. Rispetto al codice PREMAR precedentemente sviluppato, il codice
Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code
Energy Technology Data Exchange (ETDEWEB)
Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)
2015-07-01
Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)
Penelope - a code system for Monte Carlo simulation of electron and photon transport
International Nuclear Information System (INIS)
2003-01-01
Radiation is used in many applications of modern technology. Its proper handling requires competent knowledge of the basic physical laws governing its interaction with matter. To ensure its safe use, appropriate tools for predicting radiation fields and doses, as well as pertinent regulations, are required. One area of radiation physics that has received much attention concerns electron-photon transport in matter. PENELOPE is a modern, general-purpose Monte Carlo tool for simulating the transport of electrons and photons, which is applicable for arbitrary materials and in a wide energy range. PENELOPE provides quantitative guidance for many practical situations and techniques, including electron and X-ray spectroscopies, electron microscopy and microanalysis, biophysics, dosimetry, medical diagnostics and radiotherapy, as well as radiation damage and shielding. These proceedings contain the extensively revised teaching notes of the second workshop/training course on PENELOPE held in 2003, along with a detailed description of the improved physic models, numerical algorithms and structure of the code system. (author)
Monte Carlo code development in Los Alamos
International Nuclear Information System (INIS)
Carter, L.L.; Cashwell, E.D.; Everett, C.J.; Forest, C.A.; Schrandt, R.G.; Taylor, W.M.; Thompson, W.L.; Turner, G.D.
1974-01-01
The present status of Monte Carlo code development at Los Alamos Scientific Laboratory is discussed. A brief summary is given of several of the most important neutron, photon, and electron transport codes. 17 references. (U.S.)
MCNP: a general Monte Carlo code for neutron and photon transport
International Nuclear Information System (INIS)
1978-07-01
The general-purpose Monte Carlo code MCNP can be used for neutron, photon, or coupled neutron--photon transport. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-IV) are accounted for. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. Standard optional variance reduction schemes include geometry splitting and Russian roulette, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point detectors, track-length estimators, and source biasing. The standard output of MCNP includes two-way current as a function of energy, time, and angle with the normal, across any subset of bounding surfaces in the problem. Fluxes across any set of bounding surfaces are available as a function of time and energy. Similarly, the flux at designated points and the average flux in a cell (track length per unit volume) are standard tallies. Reactions such as fissions or absorptions may be obtained in a subset of geometric cells. The heating tallies give the energy deposition per starting particle. In addition, particles may be flagged when they cross specified surfaces or enter designated cells, and the contributions of these flagged particles to certain of the tallies are listed separately. All quantities printed out have their relative errors listed also. 11 figures, 27 tables
Modelling of a general purpose irradiation chamber using a Monte Carlo particle transport code
International Nuclear Information System (INIS)
Dhiyauddin Ahmad Fauzi; Sheik, F.O.A.; Nurul Fadzlin Hasbullah
2013-01-01
Full-text: The aim of this research is to stimulate the effectiveness use of a general purpose irradiation chamber to contain pure neutron particles obtained from a research reactor. The secondary neutron and gamma particles dose discharge from the chamber layers will be used as a platform to estimate the safe dimension of the chamber. The chamber, made up of layers of lead (Pb), shielding, polyethylene (PE), moderator and commercial grade aluminium (Al) cladding is proposed for the use of interacting samples with pure neutron particles in a nuclear reactor environment. The estimation was accomplished through simulation based on general Monte Carlo N-Particle transport code using Los Alamos MCNPX software. Simulations were performed on the model of the chamber subjected to high neutron flux radiation and its gamma radiation product. The model of neutron particle used is based on the neutron source found in PUSPATI TRIGA MARK II research reactor which holds a maximum flux value of 1 x 10 12 neutron/ cm 2 s. The expected outcomes of this research are zero gamma dose in the core of the chamber and neutron dose rate of less than 10 μSv/ day discharge from the chamber system. (author)
ITS Version 3.0: The Integrated TIGER Series of coupled electron/photon Monte Carlo transport codes
International Nuclear Information System (INIS)
Halbleib, J.A.; Kensek, R.P.; Valdez, G.D.; Mehlhorn, T.A.; Seltzer, S.M.; Berger, M.J.
1993-01-01
ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields. It combines operational simplicity and physical accuracy in order to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Flexibility of construction permits tailoring of the codes to specific applications and extension of code capabilities to more complex applications through simple update procedures
Monte-Carlo Impurity transport simulations in the edge of the DIII-D tokamak using the MCI code
International Nuclear Information System (INIS)
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
ITS Version 3.0: The Integrated TIGER Series of coupled electron/photon Monte Carlo transport codes
Energy Technology Data Exchange (ETDEWEB)
Halbleib, J.A.; Kensek, R.P.; Valdez, G.D.; Mehlhorn, T.A. [Sandia National Labs., Albuquerque, NM (United States); Seltzer, S.M.; Berger, M.J. [National Inst. of Standards and Technology, Gaithersburg, MD (United States). Ionizing Radiation Div.
1993-06-01
ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields. It combines operational simplicity and physical accuracy in order to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Flexibility of construction permits tailoring of the codes to specific applications and extension of code capabilities to more complex applications through simple update procedures.
Monte-Carlo Impurity transport simulations in the edge of the DIII-D tokamak using the MCI code
International Nuclear Information System (INIS)
Evans, T.E.; Sager, G.T.; Mahdavi, M.A.; Porter, G.D.; Fenstermacher, M.E.; Meyer, W.H.
1995-01-01
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 DII-D divertor is discussed. MCI simulation results are compared to experimental DII-D carbon measurements. 2 refs
Vectorization and multitasking with a Monte-Carlo code for neutron transport problems
International Nuclear Information System (INIS)
Chauvet, Y.
1985-04-01
This paper summarizes two improvements of a Monte Carlo code by resorting to vectorization and multitasking techniques. After a short presentation of the physical problem to solve and a description of the main difficulties to produce an efficient coding, this paper introduces the vectorization principles employed and briefly describes how the vectorized algorithm works. Next, measured performances on CRAY 1S, CYBER 205 and CRAY X-MP are compared. The second part of this paper is devoted to multitasking technique. Starting from the standard multitasking tools available with FORTRAN on CRAY X-MP/4, a multitasked algorithm and its measured speed-ups are presented. In conclusion we prove that vector and parallel computers are a great opportunity for such Monte Carlo algorithms
Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code
International Nuclear Information System (INIS)
Leppaenen, J.
2007-01-01
High-temperature gas-cooled reactor fuels are composed of thousands of microscopic fuel particles, randomly dispersed in a graphite matrix. The modelling of such geometry is complicated, especially using continuous-energy Monte Carlo codes, which are unable to apply any deterministic corrections in the calculation. This paper presents the geometry routine developed for modelling randomly dispersed particle fuels using the PSG Monte Carlo reactor physics code. The model is based on the delta-tracking method, and it takes into account the spatial self-shielding effects and the random dispersion of the fuel particles. The calculation routine is validated by comparing the results to reference MCNP4C calculations using uranium and plutonium based fuels. (authors)
SPHERE: a spherical-geometry multimaterial electron/photon Monte Carlo transport code
International Nuclear Information System (INIS)
Halbleib, J.A. Sr.
1977-06-01
SPHERE provides experimenters and theorists with a method for the routine solution of coupled electron/photon transport through multimaterial configurations possessing spherical symmetry. Emphasis is placed upon operational simplicity without sacrificing the rigor of the model. SPHERE combines condensed-history electron Monte Carlo with conventional single-scattering photon Monte Carlo in order to describe the transport of all generations of particles from several MeV down to 1.0 and 10.0 keV for electrons and photons, respectively. The model is more accurate at the higher energies, with a less rigorous description of the particle cascade at energies where the shell structure of the transport media becomes important. Flexibility of construction permits the user to tailor the model to specific applications and to extend the capabilities of the model to more sophisticated applications through relatively simple update procedures. 8 figs., 3 tables
Installation of Monte Carlo neutron and photon transport code system MCNP4
International Nuclear Information System (INIS)
Takano, Makoto; Sasaki, Mikio; Kaneko, Toshiyuki; Yamazaki, Takao.
1993-03-01
The continuous energy Monte Carlo code MCNP-4 including its graphic functions has been installed on the Sun-4 sparc-2 work station with minor corrections. In order to validate the installed MCNP-4 code, 25 sample problems have been executed on the work station and these results have been compared with the original ones. And, the most of the graphic functions have been demonstrated by using 3 sample problems. Further, additional 14 nuclides have been included to the continuous cross section library edited from JENDL-3. (author)
A Monte Carlo Code for Relativistic Radiation Transport Around Kerr Black Holes
Schnittman, Jeremy David; Krolik, Julian H.
2013-01-01
We present a new code for radiation transport around Kerr black holes, including arbitrary emission and absorption mechanisms, as well as electron scattering and polarization. The code is particularly useful for analyzing accretion flows made up of optically thick disks and optically thin coronae. We give a detailed description of the methods employed in the code and also present results from a number of numerical tests to assess its accuracy and convergence.
Automatic modeling for the Monte Carlo transport code Geant4 in MCAM
International Nuclear Information System (INIS)
Nie Fanzhi; Hu Liqin; Wang Guozhong; Wang Dianxi; Wu Yican; Wang Dong; Long Pengcheng; FDS Team
2014-01-01
Geant4 is a widely used Monte Carlo transport simulation package. Its geometry models could be described in geometry description markup language (GDML), but it is time-consuming and error-prone to describe the geometry models manually. This study implemented the conversion between computer-aided design (CAD) geometry models and GDML models. The conversion program was integrated into Multi-Physics Coupling Analysis Modeling Program (MCAM). The tests, including FDS-Ⅱ model, demonstrated its accuracy and feasibility. (authors)
Convergence acceleration in the Monte-Carlo particle transport code TRIPOLI-4 in criticality
International Nuclear Information System (INIS)
Dehaye, Benjamin
2014-01-01
Fields such as criticality studies need to compute some values of interest in neutron physics. Two kind of codes may be used: deterministic ones and stochastic ones. The stochastic codes do not require approximation and are thus more exact. However, they may require a lot of time to converge with a sufficient precision.The work carried out during this thesis aims to build an efficient acceleration strategy in the TRIPOLI-4. We wish to implement the zero variance game. To do so, the method requires to compute the adjoint flux. The originality of this work is to directly compute the adjoint flux directly from a Monte-Carlo simulation without using external codes thanks to the fission matrix method. This adjoint flux is then used as an importance map to bias the simulation. (author) [fr
International Nuclear Information System (INIS)
Brenner, D.J.; Prael, R.E.; Little, R.C.
1987-01-01
Realistic simulations of the passage of fast neutrons through tissue require a large quantity of cross-sectional data. What are needed are differential (in particle type, energy and angle) cross sections. A computer code is described which produces such spectra for neutrons above ∼14 MeV incident on light nuclei such as carbon and oxygen. Comparisons have been made with experimental measurements of double-differential secondary charged-particle production on carbon and oxygen at energies from 27 to 60 MeV; they indicate that the model is adequate in this energy range. In order to utilize fully the results of these calculations, they should be incorporated into a neutron transport code. This requires defining a generalized format for describing charged-particle production, putting the calculated results in this format, interfacing the neutron transport code with these data, and charged-particle transport. The design and development of such a program is described. 13 refs., 3 figs
The Premar Code for the Monte Carlo Simulation of Radiation Transport In the Atmosphere
International Nuclear Information System (INIS)
Cupini, E.; Borgia, M.G.; Premuda, M.
1997-03-01
The Montecarlo code PREMAR is described, which allows the user to simulate the radiation transport in the atmosphere, in the ultraviolet-infrared frequency interval. A plan multilayer geometry is at present foreseen by the code, witch albedo possibility at the lower boundary surface. For a given monochromatic point source, the main quantities computed by the code are the absorption spatial distributions of aerosol and molecules, together with the related atmospheric transmittances. Moreover, simulation of of Lidar experiments are foreseen by the code, the source and telescope fields of view being assigned. To build-up the appropriate probability distributions, an input data library is assumed to be read by the code. For this purpose the radiance-transmittance LOWTRAN-7 code has been conveniently adapted as a source of the library so as to exploit the richness of information of the code for a large variety of atmospheric simulations. Results of applications of the PREMAR code are finally presented, with special reference to simulations of Lidar system and radiometer experiments carried out at the Brasimone ENEA Centre by the Environment Department
International Nuclear Information System (INIS)
Wagner, John C.; Peplow, Douglas E.; Mosher, Scott W.; Evans, Thomas M.
2010-01-01
This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(10 2-4 ), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.
International Nuclear Information System (INIS)
Wagner, John C.; Peplow, Douglas E.; Mosher, Scott W.; Evans, Thomas M.
2010-01-01
This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(102-4), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.
International Nuclear Information System (INIS)
Wagner, J.C.; Peplow, D.E.; Mosher, S.W.; Evans, T.M.
2010-01-01
This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(10 2-4 ), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications. (author)
Towards scalable parellelism in Monte Carlo particle transport codes using remote memory access
Energy Technology Data Exchange (ETDEWEB)
Romano, Paul K [Los Alamos National Laboratory; Brown, Forrest B [Los Alamos National Laboratory; Forget, Benoit [MIT
2010-01-01
One forthcoming challenge in the area of high-performance computing is having the ability to run large-scale problems while coping with less memory per compute node. In this work, they investigate a novel data decomposition method that would allow Monte Carlo transport calculations to be performed on systems with limited memory per compute node. In this method, each compute node remotely retrieves a small set of geometry and cross-section data as needed and remotely accumulates local tallies when crossing the boundary of the local spatial domain. initial results demonstrate that while the method does allow large problems to be run in a memory-limited environment, achieving scalability may be difficult due to inefficiencies in the current implementation of RMA operations.
A Monte Carlo transport code study of the space radiation environment using FLUKA and ROOT
Wilson, T; Carminati, F; Brun, R; Ferrari, A; Sala, P; Empl, A; MacGibbon, J
2001-01-01
We report on the progress of a current study aimed at developing a state-of-the-art Monte-Carlo computer simulation of the space radiation environment using advanced computer software techniques recently available at CERN, the European Laboratory for Particle Physics in Geneva, Switzerland. By taking the next-generation computer software appearing at CERN and adapting it to known problems in the implementation of space exploration strategies, this research is identifying changes necessary to bring these two advanced technologies together. The radiation transport tool being developed is tailored to the problem of taking measured space radiation fluxes impinging on the geometry of any particular spacecraft or planetary habitat and simulating the evolution of that flux through an accurate model of the spacecraft material. The simulation uses the latest known results in low-energy and high-energy physics. The output is a prediction of the detailed nature of the radiation environment experienced in space as well a...
Towards scalable parallelism in Monte Carlo particle transport codes using remote memory access
International Nuclear Information System (INIS)
Romano, Paul K.; Forget, Benoit; Brown, Forrest
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, we 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. (author)
International Nuclear Information System (INIS)
Franke, B.C.; Kensek, R.P.; Prinja, A.K.
2013-01-01
Stochastic-media simulations require numerous boundary crossings. We consider two Monte Carlo electron transport approaches and evaluate accuracy with numerous material boundaries. In the condensed-history method, approximations are made based on infinite-medium solutions for multiple scattering over some track length. Typically, further approximations are employed for material-boundary crossings where infinite-medium solutions become invalid. We have previously explored an alternative 'condensed transport' formulation, a Generalized Boltzmann-Fokker-Planck (GBFP) method, which requires no special boundary treatment but instead uses approximations to the electron-scattering cross sections. Some limited capabilities for analog transport and a GBFP method have been implemented in the Integrated Tiger Series (ITS) codes. Improvements have been made to the condensed history algorithm. The performance of the ITS condensed-history and condensed-transport algorithms are assessed for material-boundary crossings. These assessments are made both by introducing artificial material boundaries and by comparison to analog Monte Carlo simulations. (authors)
Geometry system used in the General Monte Carlo transport code SPARTAN
International Nuclear Information System (INIS)
Bending, R.C.; Easter, P.G.
1974-01-01
The geometry routines used in the general-purpose, three-dimensional particle transport code SPARTAN are described. The code is designed to deal with the very complex geometries encountered in lattice cell and fuel handling calculations, health physics, and shielding problems. Regions of the system being studied may be represented by simple shapes (spheres, cylinders, and so on) or by multinomial surfaces of any order, and many simple shapes may be combined to make up a complex layout. The geometry routines are designed to allow the program to carry out a number of tasks (such as sampling for a random point or tracking a path through several regions) in any order, so that the use of the routines is not restricted to a particular tracking or scoring method. Routines for reading, checking, and printing the data are included. (U.S.)
MORSE-CGT Monte Carlo radiation transport code with the capability of the torus geometric treatment
International Nuclear Information System (INIS)
Deng Li
1990-01-01
The combinatorial geometry package CGT with the capability of the torus geometric treatment is introduced. It is get by developing the combinatorial geometry package CG. The CGT package can be transplanted to those codes which the CG package is being used and makes them also with the capability. The MORSE-CGT code can be used to solve the neutron, gamma-ray or coupled neutron-gamma-ray transport problems and time dependence for both shielding and criticality problems in torus system or system which is produced by arbitrary finite combining torus with torus or other bodies in CG package and it can also be used to design the blanket and compute shielding for TOKAMAK Fusion-Fission Hybrid Reactor
International Nuclear Information System (INIS)
Pölz, Stefan; Laubersheimer, Sven; Eberhardt, Jakob S; Harrendorf, Marco A; Keck, Thomas; Benzler, Andreas; Breustedt, Bastian
2013-01-01
The basic idea of Voxel2MCNP is to provide a framework supporting users in modeling radiation transport scenarios using voxel phantoms and other geometric models, generating corresponding input for the Monte Carlo code MCNPX, and evaluating simulation output. Applications at Karlsruhe Institute of Technology are primarily whole and partial body counter calibration and calculation of dose conversion coefficients. A new generic data model describing data related to radiation transport, including phantom and detector geometries and their properties, sources, tallies and materials, has been developed. It is modular and generally independent of the targeted Monte Carlo code. The data model has been implemented as an XML-based file format to facilitate data exchange, and integrated with Voxel2MCNP to provide a common interface for modeling, visualization, and evaluation of data. Also, extensions to allow compatibility with several file formats, such as ENSDF for nuclear structure properties and radioactive decay data, SimpleGeo for solid geometry modeling, ImageJ for voxel lattices, and MCNPX’s MCTAL for simulation results have been added. The framework is presented and discussed in this paper and example workflows for body counter calibration and calculation of dose conversion coefficients is given to illustrate its application. (paper)
International Nuclear Information System (INIS)
Masukawa, Fumihiro; Takano, Makoto; Naito, Yoshitaka; Yamazaki, Takao; Fujisaki, Masahide; Suzuki, Koichiro; Okuda, Motoi.
1993-11-01
In order to improve the accuracy and calculating speed of shielding analyses, MCNP 4, a Monte Carlo neutron and photon transport code system, has been parallelized and measured of its efficiency in the highly parallel distributed memory type computer, AP1000. The code has been analyzed statically and dynamically, then the suitable algorithm for parallelization has been determined for the shielding analysis functions of MCNP 4. This includes a strategy where a new history is assigned to the idling processor element dynamically during the execution. Furthermore, to avoid the congestion of communicative processing, the batch concept, processing multi-histories by a unit, has been introduced. By analyzing a sample cask problem with 2,000,000 histories by the AP1000 with 512 processor elements, the 82 % of parallelization efficiency is achieved, and the calculational speed has been estimated to be around 50 times as fast as that of FACOM M-780. (author)
User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code
International Nuclear Information System (INIS)
Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.
2003-01-01
This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k eff (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)
MCNP-DSP, Monte Carlo Neutron-Particle Transport Code with Digital Signal Processing
International Nuclear Information System (INIS)
2002-01-01
1 - Description of program or function: MCNP-DSP is recommended only for experienced MCNP users working with subcritical measurements. It is a modification of the Los Alamos National Laboratory's Monte Carlo code MCNP4a that is used to simulate a variety of subcritical measurements. The DSP version was developed to simulate frequency analysis measurements, correlation (Rossi-) measurements, pulsed neutron measurements, Feynman variance measurements, and multiplicity measurements. CCC-700/MCNP4C is recommended for general purpose calculations. 2 - Methods:MCNP-DSP performs calculations very similarly to MCNP and uses the same generalized geometry capabilities of MCNP. MCNP-DSP can only be used with the continuous-energy cross-section data. A variety of source and detector options are available. However, unlike standard MCNP, the source and detector options are limited to those described in the manual because these options are specified in the MCNP-DSP extra data file. MCNP-DSP is used to obtain the time-dependent response of detectors that are modeled in the simulation geometry. The detectors represent actual detectors used in measurements. These time-dependent detector responses are used to compute a variety of quantities such as frequency analysis signatures, correlation signatures, multiplicity signatures, etc., between detectors or sources and detectors. Energy ranges are 0-60 MeV for neutrons (data generally only available up to 20 MeV) and 1 keV - 1 GeV for photons and electrons. 3 - Restrictions on the complexity of the problem: None noted
International Nuclear Information System (INIS)
Günay, Mehtap; Şarer, Başar; Kasap, Hızır
2014-01-01
Highlights: • The effects of some fluids on gas production rates in structural material were investigated. • The MCNPX-2.7.0 Monte Carlo code was used for three-dimensional calculations. • It was found that biggest contribution to gas production rates comes from Fe isotope of the. • The desirable values for 5% SFG-PuO 2 with respect to radiation damage were specified. - Abstract: In this study, the molten salt-heavy metal mixtures 99–95% Li20Sn80-1-5% SFG-Pu, 99–95% Li20Sn80-1-5% SFG-PuF4, 99-95% Li20Sn80-1-5% SFG-PuO2 were used as fluids. The fluids were used in the liquid first-wall, blanket and shield zones of the designed hybrid reactor system. 9Cr2WVTa ferritic steel with the width of 4 cm was used as the structural material. The parameters of radiation damage are proton, deuterium, tritium, He-3 and He-4 gas production rates. In this study, the effects of the selected fluid on the radiation damage, in terms of individual as well as total isotopes in the structural material, were investigated for 30 full power years (FPYs). Three-dimensional analyses were performed using the most recent version of the MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library
Monte Carlo Codes Invited Session
International Nuclear Information System (INIS)
Trama, J.C.; Malvagi, F.; Brown, F.
2013-01-01
This document lists 22 Monte Carlo codes used in radiation transport applications throughout the world. For each code the names of the organization and country and/or place are given. We have the following computer codes. 1) ARCHER, USA, RPI; 2) COG11, USA, LLNL; 3) DIANE, France, CEA/DAM Bruyeres; 4) FLUKA, Italy and CERN, INFN and CERN; 5) GEANT4, International GEANT4 collaboration; 6) KENO and MONACO (SCALE), USA, ORNL; 7) MC21, USA, KAPL and Bettis; 8) MCATK, USA, LANL; 9) MCCARD, South Korea, Seoul National University; 10) MCNP6, USA, LANL; 11) MCU, Russia, Kurchatov Institute; 12) MONK and MCBEND, United Kingdom, AMEC; 13) MORET5, France, IRSN Fontenay-aux-Roses; 14) MVP2, Japan, JAEA; 15) OPENMC, USA, MIT; 16) PENELOPE, Spain, Barcelona University; 17) PHITS, Japan, JAEA; 18) PRIZMA, Russia, VNIITF; 19) RMC, China, Tsinghua University; 20) SERPENT, Finland, VTT; 21) SUPERMONTECARLO, China, CAS INEST FDS Team Hefei; and 22) TRIPOLI-4, France, CEA Saclay
Energy Technology Data Exchange (ETDEWEB)
Vergnaud, Th.; Nimal, J.C.; Chiron, M
2001-07-01
The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)
Energy Technology Data Exchange (ETDEWEB)
Vergnaud, Th; Nimal, J C; Chiron, M
2001-07-01
The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)
International Nuclear Information System (INIS)
Greenman, G.M.; O'Brien, M.J.; Procassini, R.J.; Joy, K.I.
2009-01-01
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
International Nuclear Information System (INIS)
Asano, K.; Ohno, N.; Takamura, S.
2001-01-01
Monte Carlo simulation code on impurity transport has been developed by several groups to be utilized mainly for fusion related edge plasmas. State of impurity particle is determined by atomic and molecular processes such as ionization, charge exchange in plasma. A lot of atomic and molecular processes have been considered because the edge plasma has not only impurity atoms, but also impurity molecules mainly related to chemical erosion of carbon materials, and their cross sections have been given experimentally and theoretically. We need to reveal which process is essential in a given edge plasma condition. Monte Carlo simulation code, which takes such various atomic and molecular processes into account, is necessary to investigate the behavior of impurity particle in plasmas. Usually, the impurity transport simulation code has been intended for some specific atomic and molecular processes so that the introduction of a new process forces complicated programming work. In order to evaluate various proposed atomic and molecular processes, a flexible management of atomic and molecular reaction should be established. We have developed the impurity transport simulation code based on object-oriented method. By employing object-oriented programming, we can handle each particle as 'object', which enfolds data and procedure function itself. A user (notice, not programmer) can define property of each particle species and the related atomic and molecular processes and then each 'object' is defined by analyzing this information. According to the relation among plasma particle species, objects are connected with each other and change their state by themselves. Dynamic allocation of these objects to program memory is employed to adapt for arbitrary number of species and atomic/molecular reactions. Thus we can treat arbitrary species and process starting from, for instance, methane and acetylene. Such a software procedure would be useful also for industrial application plasmas
International Nuclear Information System (INIS)
Grieshemer, D.P.; Gill, D.F.; Nease, B.R.; Carpenter, D.C.; Joo, H.; Millman, D.L.; Sutton, T.M.; Stedry, M.H.; Dobreff, P.S.; Trumbull, T.H.; Caro, E.
2013-01-01
MC21 is a continuous-energy Monte Carlo radiation transport code for the calculation of the steady-state spatial distributions of reaction rates in three-dimensional models. The code supports neutron and photon transport in fixed source problems, as well as iterated-fission-source (eigenvalue) neutron transport problems. MC21 has been designed and optimized to support large-scale problems in reactor physics, shielding, and criticality analysis applications. The code also supports many in-line reactor feedback effects, including depletion, thermal feedback, xenon feedback, eigenvalue search, and neutron and photon heating. MC21 uses continuous-energy neutron/nucleus interaction physics over the range from 10 -5 eV to 20 MeV. The code treats all common neutron scattering mechanisms, including fast-range elastic and non-elastic scattering, and thermal- and epithermal-range scattering from molecules and crystalline materials. For photon transport, MC21 uses continuous-energy interaction physics over the energy range from 1 keV to 100 GeV. The code treats all common photon interaction mechanisms, including Compton scattering, pair production, and photoelectric interactions. All of the nuclear data required by MC21 is provided by the NDEX system of codes, which extracts and processes data from EPDL-, ENDF-, and ACE-formatted source files. For geometry representation, MC21 employs a flexible constructive solid geometry system that allows users to create spatial cells from first- and second-order surfaces. The system also allows models to be built up as hierarchical collections of previously defined spatial cells, with interior detail provided by grids and template overlays. Results are collected by a generalized tally capability which allows users to edit integral flux and reaction rate information. Results can be collected over the entire problem or within specific regions of interest through the use of phase filters that control which particles are allowed to score each
Energy Technology Data Exchange (ETDEWEB)
Liu, T.; Ding, A.; Ji, W.; Xu, X. G. [Nuclear Engineering and Engineering Physics, Rensselaer Polytechnic Inst., Troy, NY 12180 (United States); Carothers, C. D. [Dept. of Computer Science, Rensselaer Polytechnic Inst. RPI (United States); Brown, F. B. [Los Alamos National Laboratory (LANL) (United States)
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)
International Nuclear Information System (INIS)
Liu, T.; Ding, A.; Ji, W.; Xu, X. G.; Carothers, C. D.; Brown, F. B.
2012-01-01
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 ∼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)
Successful vectorization - reactor physics Monte Carlo code
International Nuclear Information System (INIS)
Martin, W.R.
1989-01-01
Most particle transport Monte Carlo codes in use today are based on the ''history-based'' algorithm, wherein one particle history at a time is simulated. Unfortunately, the ''history-based'' approach (present in all Monte Carlo codes until recent years) is inherently scalar and cannot be vectorized. In particular, the history-based algorithm cannot take advantage of vector architectures, which characterize the largest and fastest computers at the current time, vector supercomputers such as the Cray X/MP or IBM 3090/600. However, substantial progress has been made in recent years in developing and implementing a vectorized Monte Carlo algorithm. This algorithm follows portions of many particle histories at the same time and forms the basis for all successful vectorized Monte Carlo codes that are in use today. This paper describes the basic vectorized algorithm along with descriptions of several variations that have been developed by different researchers for specific applications. These applications have been mainly in the areas of neutron transport in nuclear reactor and shielding analysis and photon transport in fusion plasmas. The relative merits of the various approach schemes will be discussed and the present status of known vectorization efforts will be summarized along with available timing results, including results from the successful vectorization of 3-D general geometry, continuous energy Monte Carlo. (orig.)
Monte Carlo codes use in neutron therapy
International Nuclear Information System (INIS)
Paquis, P.; Mokhtari, F.; Karamanoukian, D.; Pignol, J.P.; Cuendet, P.; Iborra, N.
1998-01-01
Monte Carlo calculation codes allow to study accurately all the parameters relevant to radiation effects, like the dose deposition or the type of microscopic interactions, through one by one particle transport simulation. These features are very useful for neutron irradiations, from device development up to dosimetry. This paper illustrates some applications of these codes in Neutron Capture Therapy and Neutron Capture Enhancement of fast neutrons irradiations. (authors)
Monte Carlo codes and Monte Carlo simulator program
International Nuclear Information System (INIS)
Higuchi, Kenji; Asai, Kiyoshi; Suganuma, Masayuki.
1990-03-01
Four typical Monte Carlo codes KENO-IV, MORSE, MCNP and VIM have been vectorized on VP-100 at Computing Center, JAERI. The problems in vector processing of Monte Carlo codes on vector processors have become clear through the work. As the result, it is recognized that these are difficulties to obtain good performance in vector processing of Monte Carlo codes. A Monte Carlo computing machine, which processes the Monte Carlo codes with high performances is being developed at our Computing Center since 1987. The concept of Monte Carlo computing machine and its performance have been investigated and estimated by using a software simulator. In this report the problems in vectorization of Monte Carlo codes, Monte Carlo pipelines proposed to mitigate these difficulties and the results of the performance estimation of the Monte Carlo computing machine by the simulator are described. (author)
International Nuclear Information System (INIS)
Kirk, B.L.
1985-12-01
The ITS (Integrated Tiger Series) Monte Carlo code package developed at Sandia National Laboratories and distributed as CCC-467/ITS by the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory (ORNL) consists of eight codes - the standard codes, TIGER, CYLTRAN, ACCEPT; the P-codes, TIGERP, CYLTRANP, ACCEPTP; and the M-codes ACCEPTM, CYLTRANM. The codes have been adapted to run on the IBM 3081, VAX 11/780, CDC-7600, and Cray 1 with the use of the update emulator UPEML. This manual should serve as a guide to a user running the codes on IBM computers having 370 architecture. The cases listed were tested on the IBM 3033, under the MVS operating system using the VS Fortran Level 1.3.1 compiler
International Nuclear Information System (INIS)
Clancy, B.E.
1986-01-01
This chapter begins with a neutron transport equation which includes the one dimensional plane geometry problems, the one dimensional spherical geometry problems, and numerical solutions. The section on the ANISN code and its look-alikes covers problems which can be solved; eigenvalue problems; outer iteration loop; inner iteration loop; and finite difference solution procedures. The input and output data for ANISN is also discussed. Two dimensional problems such as the DOT code are given. Finally, an overview of the Monte-Carlo methods and codes are elaborated on
Energy Technology Data Exchange (ETDEWEB)
Guenay, Mehtap [Inoenue Univ., Malatya (Turkey). Physics Dept.
2014-04-15
In this study, the molten salt-heavy metal mixtures 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % UO{sub 2}, 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % NpO{sub 2}, 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % UCO were used as fluids. The fluids were used in the liquid first wall, blanket and shield zones of the designed hybrid reactor system. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. In this study, the effect of mixture components on the neutron flux was investigated in a designed fusion-fission hybrid reactor system. The neutron flux was investigated according to the mixture components, radial flux distribution and energy spectrum in the designed system. Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library. (orig.)
International Nuclear Information System (INIS)
Raisali, G.R.
1992-01-01
A series of computer codes based on point kernel technique and also Monte Carlo method have been developed. These codes perform radiation transport calculations for irradiator systems having cartesian, cylindrical and mixed geometries. The monte Carlo calculations, the computer code 'EGS4' has been applied to a radiation processing type problem. This code has been acompanied by a specific user code. The set of codes developed include: GCELLS, DOSMAPM, DOSMAPC2 which simulate the radiation transport in gamma irradiator systems having cylinderical, cartesian, and mixed geometries, respectively. The program 'DOSMAP3' based on point kernel technique, has been also developed for dose rate mapping calculations in carrier type gamma irradiators. Another computer program 'CYLDETM' as a user code for EGS4 has been also developed to simulate dose variations near the interface of heterogeneous media in gamma irradiator systems. In addition a system of computer codes 'PRODMIX' has been developed which calculates the absorbed dose in the products with different densities. validation studies of the calculated results versus experimental dosimetry has been performed and good agreement has been obtained
International Nuclear Information System (INIS)
White, Morgan C.
2000-01-01
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V and V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to
Energy Technology Data Exchange (ETDEWEB)
White, Morgan C. [Univ. of Florida, Gainesville, FL (United States)
2000-07-01
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second
Monte Carlo codes use in neutron therapy; Application de codes Monte Carlo en neutrontherapie
Energy Technology Data Exchange (ETDEWEB)
Paquis, P.; Mokhtari, F.; Karamanoukian, D. [Hopital Pasteur, 06 - Nice (France); Pignol, J.P. [Hopital du Hasenrain, 68 - Mulhouse (France); Cuendet, P. [CEA Centre d' Etudes de Saclay, 91 - Gif-sur-Yvette (France). Direction des Reacteurs Nucleaires; Fares, G.; Hachem, A. [Faculte des Sciences, 06 - Nice (France); Iborra, N. [Centre Antoine-Lacassagne, 06 - Nice (France)
1998-04-01
Monte Carlo calculation codes allow to study accurately all the parameters relevant to radiation effects, like the dose deposition or the type of microscopic interactions, through one by one particle transport simulation. These features are very useful for neutron irradiations, from device development up to dosimetry. This paper illustrates some applications of these codes in Neutron Capture Therapy and Neutron Capture Enhancement of fast neutrons irradiations. (authors)
International Nuclear Information System (INIS)
Thiagu Supramaniam
2007-01-01
The aim of this research was to propose a new neutron collimator design for thermal neutron radiography facility using tangential beam port of PUSPATI TRIGA Mark II reactor, Malaysia Institute of Nuclear Technology Research (MINT). Best geometry and materials for neutron collimator were chosen in order to obtain a uniform beam with maximum thermal neutron flux, high L/ D ratio, high neutron to gamma ratio and low beam divergence with high resolution. Monte Carlo N-particle Transport Code version 5 (MCNP 5) was used to optimize six neutron collimator components such as beam port medium, neutron scatterer, neutron moderator, gamma filter, aperture and collimator wall. The reactor and tangential beam port setup in MCNP5 was plotted according to its actual sizes. A homogeneous reactor core was assumed and population control method of variance reduction technique was applied by using cell importance. The comparison between experimental results and simulated results of the thermal neutron flux measurement of the bare tangential beam port, shows that both graph obtained had similar pattern. This directly suggests the reliability of MCNP5 in order to obtained optimal neutron collimator parameters. The simulated results of the optimal neutron medium, shows that vacuum was the best medium to transport neutrons followed by helium gas and air. The optimized aperture component was boral with 3 cm thickness. The optimal aperture center hole diameter was 2 cm which produces 88 L/ D ratio. Simulation also shows that graphite neutron scatterer improves thermal neutron flux while reducing fast neutron flux. Neutron moderator was used to moderate fast and epithermal neutrons in the beam port. Paraffin wax with 90 cm thick was bound to be the best neutron moderator material which produces the highest thermal neutron flux at the image plane. Cylindrical shape high density polyethylene neutron collimator produces the highest thermal neutron flux at the image plane rather than divergent
Igo - A Monte Carlo Code For Radiotherapy Planning
International Nuclear Information System (INIS)
Goldstein, M.; Regev, D.
1999-01-01
The goal of radiation therapy is to deliver a lethal dose to the tumor, while minimizing the dose to normal tissues and vital organs. To carry out this task, it is critical to calculate correctly the 3-D dose delivered. Monte Carlo transport methods (especially the Adjoint Monte Carlo have the potential to provide more accurate predictions of the 3-D dose the currently used methods. IG0 is a Monte Carlo code derived from the general Monte Carlo Program - MCNP, tailored specifically for calculating the effects of radiation therapy. This paper describes the IG0 transport code, the PIG0 interface and some preliminary results
International Nuclear Information System (INIS)
Moore, J.G.
1974-01-01
The Monte Carlo code MONK is a general program written to provide a high degree of flexibility to the user. MONK is distinguished by its detailed representation of nuclear data in point form i.e., the cross-section is tabulated at specific energies instead of the more usual group representation. The nuclear data are unadjusted in the point form but recently the code has been modified to accept adjusted group data as used in fast and thermal reactor applications. The various geometrical handling capabilities and importance sampling techniques are described. In addition to the nuclear data aspects, the following features are also described; geometrical handling routines, tracking cycles, neutron source and output facilities. 12 references. (U.S.)
International Nuclear Information System (INIS)
Cai, Li
2014-01-01
calculation solver SNATCH in the PARIS code platform. The latter uses the transport theory which is indispensable for the new generation fast reactors analysis. The principal conclusions are as follows: The Monte-Carlo assembly calculation code is an interesting way (in the sense of avoiding the difficulties in the self-shielding calculation, the limited order development of anisotropy parameters, the exact 3D geometries) to validate the deterministic codes like ECCO or APOLLO3 and to produce the multi-group constants for deterministic or Monte-Carlo multi-group calculation codes. The results obtained for the moment with the multi-group constants calculated by TRIPOLI-4 code are comparable with those produced from ECCO, but did not show remarkable advantages. (author) [fr
International Nuclear Information System (INIS)
Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa
2017-03-01
In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)
International Nuclear Information System (INIS)
Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa
2017-03-01
In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)
International Nuclear Information System (INIS)
Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki
2005-06-01
In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector super-computers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, (5) continuous-energy calculation at arbitrary temperatures, (6) estimation of real variances in eigenvalue problems, (7) point detector and surface crossing estimators, (8) statistical geometry model, (9) function of reactor noise analysis (simulation of the Feynman-α experiment), (10) arbitrary shaped lattice boundary, (11) periodic boundary condition, (12) parallelization with standard libraries (MPI, PVM), (13) supporting many platforms, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them. (author)
SERPENT Monte Carlo reactor physics code
International Nuclear Information System (INIS)
Leppaenen, J.
2010-01-01
SERPENT is a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, developed at VTT Technical Research Centre of Finland since 2004. The code is specialized in lattice physics applications, but the universe-based geometry description allows transport simulation to be carried out in complicated three-dimensional geometries as well. The suggested applications of SERPENT include generation of homogenized multi-group constants for deterministic reactor simulator calculations, fuel cycle studies involving detailed assembly-level burnup calculations, validation of deterministic lattice transport codes, research reactor applications, educational purposes and demonstration of reactor physics phenomena. The Serpent code has been publicly distributed by the OECD/NEA Data Bank since May 2009 and RSICC in the U. S. since March 2010. The code is being used in some 35 organizations in 20 countries around the world. This paper presents an overview of the methods and capabilities of the Serpent code, with examples in the modelling of WWER-440 reactor physics. (Author)
Monte Carlo electron/photon transport
International Nuclear Information System (INIS)
Mack, J.M.; Morel, J.E.; Hughes, H.G.
1985-01-01
A review of nonplasma coupled electron/photon transport using Monte Carlo method is presented. Remarks are mainly restricted to linerarized formalisms at electron energies from 1 keV to 1000 MeV. Applications involving pulse-height estimation, transport in external magnetic fields, and optical Cerenkov production are discussed to underscore the importance of this branch of computational physics. Advances in electron multigroup cross-section generation is reported, and its impact on future code development assessed. Progress toward the transformation of MCNP into a generalized neutral/charged-particle Monte Carlo code is described. 48 refs
Energy Technology Data Exchange (ETDEWEB)
Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William
2005-09-01
ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 5.0, the latest version of ITS, contains (1) improvements to the ITS 3.0 continuous-energy codes, (2) multigroup codes with adjoint transport capabilities, (3) parallel implementations of all ITS codes, (4) a general purpose geometry engine for linking with CAD or other geometry formats, and (5) the Cholla facet geometry library. Moreover, the general user friendliness of the software has been enhanced through increased internal error checking and improved code portability.
A Monte Carlo burnup code linking MCNP and REBUS
International Nuclear Information System (INIS)
Hanan, N.A.; Olson, A.P.; Pond, R.B.; Matos, J.E.
1998-01-01
The REBUS-3 burnup code, used in the anl RERTR Program, is a very general code that uses diffusion theory (DIF3D) to obtain the fluxes required for reactor burnup analyses. Diffusion theory works well for most reactors. However, to include the effects of exact geometry and strong absorbers that are difficult to model using diffusion theory, a Monte Carlo method is required. MCNP, a general-purpose, generalized-geometry, time-dependent, Monte Carlo transport code, is the most widely used Monte Carlo code. This paper presents a linking of the MCNP code and the REBUS burnup code to perform these difficult analyses. The linked code will permit the use of the full capabilities of REBUS which include non-equilibrium and equilibrium burnup analyses. Results of burnup analyses using this new linked code are also presented. (author)
A Monte Carlo burnup code linking MCNP and REBUS
International Nuclear Information System (INIS)
Hanan, N. A.
1998-01-01
The REBUS-3 burnup code, used in the ANL RERTR Program, is a very general code that uses diffusion theory (DIF3D) to obtain the fluxes required for reactor burnup analyses. Diffusion theory works well for most reactors. However, to include the effects of exact geometry and strong absorbers that are difficult to model using diffusion theory, a Monte Carlo method is required. MCNP, a general-purpose, generalized-geometry, time-dependent, Monte Carlo transport code, is the most widely used Monte Carlo code. This paper presents a linking of the MCNP code and the REBUS burnup code to perform these difficult burnup analyses. The linked code will permit the use of the full capabilities of REBUS which include non-equilibrium and equilibrium burnup analyses. Results of burnup analyses using this new linked code are also presented
Verbeke, Jérôme M.; Petit, Odile; Chebboubi, Abdelhazize; Litaize, Olivier
2018-01-01
Fission modeling in general-purpose Monte Carlo transport codes often relies on average nuclear data provided by international evaluation libraries. As such, only average fission multiplicities are available and correlations between fission neutrons and photons are missing. Whereas uncorrelated fission physics is usually sufficient for standard reactor core and radiation shielding calculations, correlated fission secondaries are required for specialized nuclear instrumentation and detector modeling. For coincidence counting detector optimization for instance, precise simulation of fission neutrons and photons that remain correlated in time from birth to detection is essential. New developments were recently integrated into the Monte Carlo transport code TRIPOLI-4 to model fission physics more precisely, the purpose being to access event-by-event fission events from two different fission models: FREYA and FIFRELIN. TRIPOLI-4 simulations can now be performed, either by connecting via an API to the LLNL fission library including FREYA, or by reading external fission event data files produced by FIFRELIN beforehand. These new capabilities enable us to easily compare results from Monte Carlo transport calculations using the two fission models in a nuclear instrumentation application. In the first part of this paper, broad underlying principles of the two fission models are recalled. We then present experimental measurements of neutron angular correlations for 252Cf(sf) and 240Pu(sf). The correlations were measured for several neutron kinetic energy thresholds. In the latter part of the paper, simulation results are compared to experimental data. Spontaneous fissions in 252Cf and 240Pu are modeled by FREYA or FIFRELIN. Emitted neutrons and photons are subsequently transported to an array of scintillators by TRIPOLI-4 in analog mode to preserve their correlations. Angular correlations between fission neutrons obtained independently from these TRIPOLI-4 simulations, using
International Nuclear Information System (INIS)
Mosleh-Shirazi, Mohammad Amin; Karbasi, Sareh; Zarrini-Monfared, Zinat; Zamani, Ali
2014-01-01
Two-dimensional (2D) arrays of thick segmented scintillators are of interest as X-ray detectors for both 2D and 3D image-guided radiotherapy (IGRT). Their detection process involves ionizing radiation energy deposition followed by production and transport of optical photons. Only a very limited number of optical Monte Carlo simulation models exist, which has limited the number of modeling studies that have considered both stages of the detection process. We present ScintSim1, an in-house optical Monte Carlo simulation code for 2D arrays of scintillation crystals, developed in the MATLAB programming environment. The code was rewritten and revised based on an existing program for single-element detectors, with the additional capability to model 2D arrays of elements with configurable dimensions, material, etc., The code generates and follows each optical photon history through the detector element (and, in case of cross-talk, the surrounding ones) until it reaches a configurable receptor, or is attenuated. The new model was verified by testing against relevant theoretically known behaviors or quantities and the results of a validated single-element model. For both sets of comparisons, the discrepancies in the calculated quantities were all <1%. The results validate the accuracy of the new code, which is a useful tool in scintillation detector optimization. (author)
International Nuclear Information System (INIS)
Lee, Ki Bog; Kim, Yeong Il; Kim, Kang Seok; Kim, Sang Ji; Kim, Young Gyun; Song, Hoon; Lee, Dong Uk; Lee, Byoung Oon; Jang, Jin Wook; Lim, Hyun Jin; Kim, Hak Sung
2004-05-01
In this report, the results of KALIMER (Korea Advanced LIquid MEtal Reactor) core design calculated by the K-CORE computing system are compared and analyzed with those of MCDEP calculation. The effective multiplication factor, flux distribution, fission power distribution and the number densities of the important nuclides effected from the depletion calculation for the R-Z model and Hex-Z model of KALIMER core are compared. It is confirmed that the results of K-CORE system compared with those of MCDEP based on the Monte Carlo transport theory method agree well within 700 pcm for the effective multiplication factor estimation and also within 2% in the driver fuel region, within 10% in the radial blanket region for the reaction rate and the fission power density. Thus, the K-CORE system for the core design of KALIMER by treating the lumped fission product and mainly important nuclides can be used as a core design tool keeping the necessary accuracy
Energy Technology Data Exchange (ETDEWEB)
Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B
2003-07-01
This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k{sub eff} (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)
Electron transport code theoretical basis
International Nuclear Information System (INIS)
Dubi, A.; Horowitz, Y.S.
1978-04-01
This report mainly describes the physical and mathematical considerations involved in the treatment of the multiple collision processes. A brief description is given of the traditional methods used in electron transport via Monte Carlo, and a somewhat more detailed description, of the approach to be used in the presently developed code
Monte Carlo simulation code modernization
CERN. Geneva
2015-01-01
The continual development of sophisticated transport simulation algorithms allows increasingly accurate description of the effect of the passage of particles through matter. This modelling capability finds applications in a large spectrum of fields from medicine to astrophysics, and of course HEP. These new capabilities however come at the cost of a greater computational intensity of the new models, which has the effect of increasing the demands of computing resources. This is particularly true for HEP, where the demand for more simulation are driven by the need of both more accuracy and more precision, i.e. better models and more events. Usually HEP has relied on the "Moore's law" evolution, but since almost ten years the increase in clock speed has withered and computing capacity comes in the form of hardware architectures of many-core or accelerated processors. To harness these opportunities we need to adapt our code to concurrent programming models taking advantages of both SIMD and SIMT architectures. Th...
Rodriguez, M.; Brualla, L.
2018-04-01
Monte Carlo simulation of radiation transport is computationally demanding to obtain reasonably low statistical uncertainties of the estimated quantities. Therefore, it can benefit in a large extent from high-performance computing. This work is aimed at assessing the performance of the first generation of the many-integrated core architecture (MIC) Xeon Phi coprocessor with respect to that of a CPU consisting of a double 12-core Xeon processor in Monte Carlo simulation of coupled electron-photonshowers. The comparison was made twofold, first, through a suite of basic tests including parallel versions of the random number generators Mersenne Twister and a modified implementation of RANECU. These tests were addressed to establish a baseline comparison between both devices. Secondly, through the p DPM code developed in this work. p DPM is a parallel version of the Dose Planning Method (DPM) program for fast Monte Carlo simulation of radiation transport in voxelized geometries. A variety of techniques addressed to obtain a large scalability on the Xeon Phi were implemented in p DPM. Maximum scalabilities of 84 . 2 × and 107 . 5 × were obtained in the Xeon Phi for simulations of electron and photon beams, respectively. Nevertheless, in none of the tests involving radiation transport the Xeon Phi performed better than the CPU. The disadvantage of the Xeon Phi with respect to the CPU owes to the low performance of the single core of the former. A single core of the Xeon Phi was more than 10 times less efficient than a single core of the CPU for all radiation transport simulations.
Monte Carlo Transport for Electron Thermal Transport
Chenhall, Jeffrey; Cao, Duc; Moses, Gregory
2015-11-01
The iSNB (implicit Schurtz Nicolai Busquet multigroup electron thermal transport method of Cao et al. is adapted into a Monte Carlo transport method in order to better model the effects of non-local behavior. The end goal is a hybrid transport-diffusion method that combines Monte Carlo Transport with a discrete diffusion Monte Carlo (DDMC). The hybrid method will combine the efficiency of a diffusion method in short mean free path regions with the accuracy of a transport method in long mean free path regions. The Monte Carlo nature of the approach allows the algorithm to be massively parallelized. Work to date on the method will be presented. This work was supported by Sandia National Laboratory - Albuquerque and the University of Rochester Laboratory for Laser Energetics.
SU-E-T-180: Fano Cavity Test of Proton Transport in Monte Carlo Codes Running On GPU and Xeon Phi
International Nuclear Information System (INIS)
Sterpin, E; Sorriaux, J; Souris, K; Lee, J; Vynckier, S; Schuemann, J; Paganetti, H; Jia, X; Jiang, S
2014-01-01
Purpose: In proton dose calculation, clinically compatible speeds are now achieved with Monte Carlo codes (MC) that combine 1) adequate simplifications in the physics of transport and 2) the use of hardware architectures enabling massive parallel computing (like GPUs). However, the uncertainties related to the transport algorithms used in these codes must be kept minimal. Such algorithms can be checked with the so-called “Fano cavity test”. We implemented the test in two codes that run on specific hardware: gPMC on an nVidia GPU and MCsquare on an Intel Xeon Phi (60 cores). Methods: gPMC and MCsquare are designed for transporting protons in CT geometries. Both codes use the method of fictitious interaction to sample the step-length for each transport step. The considered geometry is a water cavity (2×2×0.2 cm 3 , 0.001 g/cm 3 ) in a 10×10×50 cm 3 water phantom (1 g/cm 3 ). CPE in the cavity is established by generating protons over the phantom volume with a uniform momentum (energy E) and a uniform intensity per unit mass I. Assuming no nuclear reactions and no generation of other secondaries, the computed cavity dose should equal IE, according to Fano's theorem. Both codes were tested for initial proton energies of 50, 100, and 200 MeV. Results: For all energies, gPMC and MCsquare are within 0.3 and 0.2 % of the theoretical value IE, respectively (0.1% standard deviation). Single-precision computations (instead of double) increased the error by about 0.1% in MCsquare. Conclusion: Despite the simplifications in the physics of transport, both gPMC and MCsquare successfully pass the Fano test. This ensures optimal accuracy of the codes for clinical applications within the uncertainties on the underlying physical models. It also opens the path to other applications of these codes, like the simulation of ion chamber response
SU-E-T-180: Fano Cavity Test of Proton Transport in Monte Carlo Codes Running On GPU and Xeon Phi
Energy Technology Data Exchange (ETDEWEB)
Sterpin, E; Sorriaux, J; Souris, K; Lee, J; Vynckier, S [Universite catholique de Louvain, Brussels, Brussels (Belgium); Schuemann, J; Paganetti, H [Massachusetts General Hospital, Boston, MA (United States); Jia, X; Jiang, S [The University of Texas Southwestern Medical Ctr, Dallas, TX (United States)
2014-06-01
Purpose: In proton dose calculation, clinically compatible speeds are now achieved with Monte Carlo codes (MC) that combine 1) adequate simplifications in the physics of transport and 2) the use of hardware architectures enabling massive parallel computing (like GPUs). However, the uncertainties related to the transport algorithms used in these codes must be kept minimal. Such algorithms can be checked with the so-called “Fano cavity test”. We implemented the test in two codes that run on specific hardware: gPMC on an nVidia GPU and MCsquare on an Intel Xeon Phi (60 cores). Methods: gPMC and MCsquare are designed for transporting protons in CT geometries. Both codes use the method of fictitious interaction to sample the step-length for each transport step. The considered geometry is a water cavity (2×2×0.2 cm{sup 3}, 0.001 g/cm{sup 3}) in a 10×10×50 cm{sup 3} water phantom (1 g/cm{sup 3}). CPE in the cavity is established by generating protons over the phantom volume with a uniform momentum (energy E) and a uniform intensity per unit mass I. Assuming no nuclear reactions and no generation of other secondaries, the computed cavity dose should equal IE, according to Fano's theorem. Both codes were tested for initial proton energies of 50, 100, and 200 MeV. Results: For all energies, gPMC and MCsquare are within 0.3 and 0.2 % of the theoretical value IE, respectively (0.1% standard deviation). Single-precision computations (instead of double) increased the error by about 0.1% in MCsquare. Conclusion: Despite the simplifications in the physics of transport, both gPMC and MCsquare successfully pass the Fano test. This ensures optimal accuracy of the codes for clinical applications within the uncertainties on the underlying physical models. It also opens the path to other applications of these codes, like the simulation of ion chamber response.
Monte Carlo code for neutron radiography
International Nuclear Information System (INIS)
Milczarek, Jacek J.; Trzcinski, Andrzej; El-Ghany El Abd, Abd; Czachor, Andrzej
2005-01-01
The concise Monte Carlo code, MSX, for simulation of neutron radiography images of non-uniform objects is presented. The possibility of modeling the images of objects with continuous spatial distribution of specific isotopes is included. The code can be used for assessment of the scattered neutron component in neutron radiograms
Monte Carlo code for neutron radiography
Energy Technology Data Exchange (ETDEWEB)
Milczarek, Jacek J. [Institute of Atomic Energy, Swierk, 05-400 Otwock (Poland)]. E-mail: jjmilcz@cyf.gov.pl; Trzcinski, Andrzej [Institute for Nuclear Studies, Swierk, 05-400 Otwock (Poland); El-Ghany El Abd, Abd [Institute of Atomic Energy, Swierk, 05-400 Otwock (Poland); Nuclear Research Center, PC 13759, Cairo (Egypt); Czachor, Andrzej [Institute of Atomic Energy, Swierk, 05-400 Otwock (Poland)
2005-04-21
The concise Monte Carlo code, MSX, for simulation of neutron radiography images of non-uniform objects is presented. The possibility of modeling the images of objects with continuous spatial distribution of specific isotopes is included. The code can be used for assessment of the scattered neutron component in neutron radiograms.
VIM: a continuous energy Monte Carlo code at ANL
International Nuclear Information System (INIS)
Blomquist, R.N.; Lell, R.M.; Gelbard, E.M.
1980-01-01
The continuous-energy Monte Carlo neutron transport code VIM and its auxiliaries are briefly described. The ENDF/B cross section data processing procedure is summarized and its benchmarking against MC 2 -2 is reviewed. Several representative applications at ANL are described, including fast critical assembly benchmark calculations and STF and TREAT Upgrade benchmark calculations. 2 figures
SPQR: a Monte Carlo reactor kinetics code
International Nuclear Information System (INIS)
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
Wareing, Todd A.; Failla, Gregory; Horton, John L.; Eifel, Patricia J.; Mourtada, Firas
2009-01-01
A patient dose distribution was calculated by a 3D multi‐group SN particle transport code for intracavitary brachytherapy of the cervix uteri and compared to previously published Monte Carlo results. A Cs‐137 LDR intracavitary brachytherapy CT data set was chosen from our clinical database. MCNPX version 2.5.c, was used to calculate the dose distribution. A 3D multi‐group SN particle transport code, Attila version 6.1.1 was used to simulate the same patient. Each patient applicator was built in SolidWorks, a mechanical design package, and then assembled with a coordinate transformation and rotation for the patient. The SolidWorks exported applicator geometry was imported into Attila for calculation. Dose matrices were overlaid on the patient CT data set. Dose volume histograms and point doses were compared. The MCNPX calculation required 14.8 hours, whereas the Attila calculation required 22.2 minutes on a 1.8 GHz AMD Opteron CPU. Agreement between Attila and MCNPX dose calculations at the ICRU 38 points was within ±3%. Calculated doses to the 2 cc and 5 cc volumes of highest dose differed by not more than ±1.1% between the two codes. Dose and DVH overlays agreed well qualitatively. Attila can calculate dose accurately and efficiently for this Cs‐137 CT‐based patient geometry. Our data showed that a three‐group cross‐section set is adequate for Cs‐137 computations. Future work is aimed at implementing an optimized version of Attila for radiotherapy calculations. PACS number: 87.53.Jw
Coded aperture optimization using Monte Carlo simulations
International Nuclear Information System (INIS)
Martineau, A.; Rocchisani, J.M.; Moretti, J.L.
2010-01-01
Coded apertures using Uniformly Redundant Arrays (URA) have been unsuccessfully evaluated for two-dimensional and three-dimensional imaging in Nuclear Medicine. The images reconstructed from coded projections contain artifacts and suffer from poor spatial resolution in the longitudinal direction. We introduce a Maximum-Likelihood Expectation-Maximization (MLEM) algorithm for three-dimensional coded aperture imaging which uses a projection matrix calculated by Monte Carlo simulations. The aim of the algorithm is to reduce artifacts and improve the three-dimensional spatial resolution in the reconstructed images. Firstly, we present the validation of GATE (Geant4 Application for Emission Tomography) for Monte Carlo simulations of a coded mask installed on a clinical gamma camera. The coded mask modelling was validated by comparison between experimental and simulated data in terms of energy spectra, sensitivity and spatial resolution. In the second part of the study, we use the validated model to calculate the projection matrix with Monte Carlo simulations. A three-dimensional thyroid phantom study was performed to compare the performance of the three-dimensional MLEM reconstruction with conventional correlation method. The results indicate that the artifacts are reduced and three-dimensional spatial resolution is improved with the Monte Carlo-based MLEM reconstruction.
International Nuclear Information System (INIS)
Iwamoto, Y.; Ogawa, T.
2016-01-01
The modelling of the damage in materials irradiated by neutrons is needed for understanding the mechanism of radiation damage in fission and fusion reactor facilities. The molecular dynamics simulations of damage cascades with full atomic interactions require information about the energy distribution of the Primary Knock on Atoms (PKAs). The most common process to calculate PKA energy spectra under low-energy neutron irradiation is to use the nuclear data processing code NJOY2012. It calculates group-to-group recoil cross section matrices using nuclear data libraries in ENDF data format, which is energy and angular recoil distributions for many reactions. After the NJOY2012 process, SPKA6C is employed to produce PKA energy spectra combining recoil cross section matrices with an incident neutron energy spectrum. However, intercomparison with different processes and nuclear data libraries has not been studied yet. Especially, the higher energy (~5 MeV) of the incident neutrons, compared to fission, leads to many reaction channels, which produces a complex distribution of PKAs in energy and type. Recently, we have developed the event generator mode (EGM) in the Particle and Heavy Ion Transport code System PHITS for neutron incident reactions in the energy region below 20 MeV. The main feature of EGM is to produce PKA with keeping energy and momentum conservation in a reaction. It is used for event-by-event analysis in application fields such as soft error analysis in semiconductors, micro dosimetry in human body, and estimation of Displacement per Atoms (DPA) value in metals and so on. The purpose of this work is to specify differences of PKA spectra and heating number related with kerma between different calculation method using PHITS-EGM and NJOY2012+SPKA6C with different libraries TENDL-2015, ENDF/B-VII.1 and JENDL-4.0 for fusion relevant materials
Energy Technology Data Exchange (ETDEWEB)
Iwamoto, Yosuke, E-mail: iwamoto.yosuke@jaea.go.jp; Ogawa, Tatsuhiko
2017-04-01
Because primary knock-on atoms (PKAs) create point defects and clusters in materials that are irradiated with neutrons, it is important to validate the calculations of recoil cross section spectra that are used to estimate radiation damage in materials. Here, the recoil cross section spectra of fission- and fusion-relevant materials were calculated using the Event Generator Mode (EGM) of the Particle and Heavy Ion Transport code System (PHITS) and also using the data processing code NJOY2012 with the nuclear data libraries TENDL2015, ENDF/BVII.1, and JEFF3.2. The heating number, which is the integral of the recoil cross section spectra, was also calculated using PHITS-EGM and compared with data extracted from the ACE files of TENDL2015, ENDF/BVII.1, and JENDL4.0. In general, only a small difference was found between the PKA spectra of PHITS + TENDL2015 and NJOY + TENDL2015. From analyzing the recoil cross section spectra extracted from the nuclear data libraries using NJOY2012, we found that the recoil cross section spectra were incorrect for {sup 72}Ge, {sup 75}As, {sup 89}Y, and {sup 109}Ag in the ENDF/B-VII.1 library, and for {sup 90}Zr and {sup 55}Mn in the JEFF3.2 library. From analyzing the heating number, we found that the data extracted from the ACE file of TENDL2015 for all nuclides were problematic in the neutron capture region because of incorrect data regarding the emitted gamma energy. However, PHITS + TENDL2015 can calculate PKA spectra and heating numbers correctly.
Considerations of MCNP Monte Carlo code to be used as a radiotherapy treatment planning tool.
Juste, B; Miro, R; Gallardo, S; Verdu, G; Santos, A
2005-01-01
The present work has simulated the photon and electron transport in a Theratron 780® (MDS Nordion)60Co radiotherapy unit, using the Monte Carlo transport code, MCNP (Monte Carlo N-Particle). This project explains mainly the different methodologies carried out to speedup calculations in order to apply this code efficiently in radiotherapy treatment planning.
Specialized Monte Carlo codes versus general-purpose Monte Carlo codes
International Nuclear Information System (INIS)
Moskvin, Vadim; DesRosiers, Colleen; Papiez, Lech; Lu, Xiaoyi
2002-01-01
The possibilities of Monte Carlo modeling for dose calculations and optimization treatment are quite limited in radiation oncology applications. The main reason is that the Monte Carlo technique for dose calculations is time consuming while treatment planning may require hundreds of possible cases of dose simulations to be evaluated for dose optimization. The second reason is that general-purpose codes widely used in practice, require an experienced user to customize them for calculations. This paper discusses the concept of Monte Carlo code design that can avoid the main problems that are preventing wide spread use of this simulation technique in medical physics. (authors)
Vectorization of Monte Carlo particle transport
International Nuclear Information System (INIS)
Burns, P.J.; Christon, M.; Schweitzer, R.; Lubeck, O.M.; Wasserman, H.J.; Simmons, M.L.; Pryor, D.V.
1989-01-01
This paper reports that fully vectorized versions of the Los Alamos National Laboratory benchmark code Gamteb, a Monte Carlo photon transport algorithm, were developed for the Cyber 205/ETA-10 and Cray X-MP/Y-MP architectures. Single-processor performance measurements of the vector and scalar implementations were modeled in a modified Amdahl's Law that accounts for additional data motion in the vector code. The performance and implementation strategy of the vector codes are related to architectural features of each machine. Speedups between fifteen and eighteen for Cyber 205/ETA-10 architectures, and about nine for CRAY X-MP/Y-MP architectures are observed. The best single processor execution time for the problem was 0.33 seconds on the ETA-10G, and 0.42 seconds on the CRAY Y-MP
General purpose code for Monte Carlo simulations
International Nuclear Information System (INIS)
Wilcke, W.W.
1983-01-01
A general-purpose computer called MONTHY has been written to perform Monte Carlo simulations of physical systems. To achieve a high degree of flexibility the code is organized like a general purpose computer, operating on a vector describing the time dependent state of the system under simulation. The instruction set of the computer is defined by the user and is therefore adaptable to the particular problem studied. The organization of MONTHY allows iterative and conditional execution of operations
The vector and parallel processing of MORSE code on Monte Carlo Machine
International Nuclear Information System (INIS)
Hasegawa, Yukihiro; Higuchi, Kenji.
1995-11-01
Multi-group Monte Carlo Code for particle transport, MORSE is modified for high performance computing on Monte Carlo Machine Monte-4. The method and the results are described. Monte-4 was specially developed to realize high performance computing of Monte Carlo codes for particle transport, which have been difficult to obtain high performance in vector processing on conventional vector processors. Monte-4 has four vector processor units with the special hardware called Monte Carlo pipelines. The vectorization and parallelization of MORSE code and the performance evaluation on Monte-4 are described. (author)
NASA space radiation transport code development consortium
International Nuclear Information System (INIS)
Townsend, L. W.
2005-01-01
Recently, NASA established a consortium involving the Univ. of Tennessee (lead institution), the Univ. of Houston, Roanoke College and various government and national laboratories, to accelerate the development of a standard set of radiation transport computer codes for NASA human exploration applications. This effort involves further improvements of the Monte Carlo codes HETC and FLUKA and the deterministic code HZETRN, including developing nuclear reaction databases necessary to extend the Monte Carlo codes to carry out heavy ion transport, and extending HZETRN to three dimensions. The improved codes will be validated by comparing predictions with measured laboratory transport data, provided by an experimental measurements consortium, and measurements in the upper atmosphere on the balloon-borne Deep Space Test Bed (DSTB). In this paper, we present an overview of the consortium members and the current status and future plans of consortium efforts to meet the research goals and objectives of this extensive undertaking. (authors)
Monte Carlo burnup codes acceleration using the correlated sampling method
International Nuclear Information System (INIS)
Dieudonne, C.
2013-01-01
For several years, Monte Carlo burnup/depletion codes have appeared, which couple Monte Carlo codes to simulate the neutron transport to deterministic methods, which handle 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 Monte Carlo solver called at each time step. In this document we present an original methodology to avoid the repetitive and time-expensive Monte Carlo simulations, and to replace them by perturbation calculations: indeed the different burnup steps may be seen as perturbations of the isotopic concentration of an initial Monte Carlo simulation. In a first time we will present this method, and provide details on the perturbative technique used, namely the correlated sampling. In a second time we develop a theoretical model to study the features of the correlated sampling method to understand its effects on depletion calculations. In a third time the implementation of this method in the TRIPOLI-4 code will be discussed, as well as the precise calculation scheme used to bring important speed-up of the depletion calculation. We will begin to validate and optimize the perturbed depletion scheme with the calculation of a REP-like fuel cell depletion. Then this technique will be used to calculate the depletion of a REP-like assembly, studied at beginning of its cycle. After having validated the method with a reference calculation we will show that it can speed-up by nearly an order of magnitude standard Monte-Carlo depletion codes. (author) [fr
Monte Carlo method for neutron transport problems
International Nuclear Information System (INIS)
Asaoka, Takumi
1977-01-01
Some methods for decreasing variances in Monte Carlo neutron transport calculations are presented together with the results of sample calculations. A general purpose neutron transport Monte Carlo code ''MORSE'' was used for the purpose. The first method discussed in this report is the method of statistical estimation. As an example of this method, the application of the coarse-mesh rebalance acceleration method to the criticality calculation of a cylindrical fast reactor is presented. Effective multiplication factor and its standard deviation are presented as a function of the number of histories and comparisons are made between the coarse-mesh rebalance method and the standard method. Five-group neutron fluxes at core center are also compared with the result of S4 calculation. The second method is the method of correlated sampling. This method was applied to the perturbation calculation of control rod worths in a fast critical assembly (FCA-V-3) Two methods of sampling (similar flight paths and identical flight paths) are tested and compared with experimental results. For every cases the experimental value lies within the standard deviation of the Monte Carlo calculations. The third method is the importance sampling. In this report a biased selection of particle flight directions discussed. This method was applied to the flux calculation in a spherical fast neutron system surrounded by a 10.16 cm iron reflector. Result-direction biasing, path-length stretching, and no biasing are compared with S8 calculation. (Aoki, K.)
Improved diffusion coefficients generated from Monte Carlo codes
International Nuclear Information System (INIS)
Herman, B. R.; Forget, B.; Smith, K.; Aviles, B. N.
2013-01-01
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)
Practical Application of Monte Carlo Code in RTP
International Nuclear Information System (INIS)
Mohamad Hairie Rabir; Julia Abdul Karim; Muhammad Rawi Mohamed Zin; Na'im Syauqi Hamzah; Mark Dennis Anak Usang; Abi Muttaqin Jalal Bayar; Muhammad Khairul Ariff Mustafa
2015-01-01
Monte Carlo neutron transport codes are widely used in various reactor physics applications in RTP and other related nuclear and radiation research in Nuklear Malaysia. The main advantage of the method is the capability to model geometry and interaction physics without major approximations. The disadvantage is that the modelling of complicated systems is very computing-intensive, which restricts the applications to some extent. The importance of Monte Carlo calculation is likely to increase in the future, along with the development in computer capacities and parallel calculation. This paper presents several calculation activities, its achievements and challenges in using MCNP code for neutronics analysis, nuclide inventory and source term calculation, shielding and dose evaluation. (author)
A PC version of the Monte Carlo criticality code OMEGA
International Nuclear Information System (INIS)
Seifert, E.
1996-05-01
A description of the PC version of the Monte Carlo criticality code OMEGA is given. The report contains a general description of the code together with a detailed input description. Furthermore, some examples are given illustrating the generation of an input file. The main field of application is the calculation of the criticality of arrangements of fissionable material. Geometrically complicated arrangements that often appear inside and outside a reactor, e.g. in a fuel storage or transport container, can be considered essentially without geometrical approximations. For example, the real geometry of assemblies containing hexagonal or square lattice structures can be described in full detail. Moreover, the code can be used for special investigations in the field of reactor physics and neutron transport. Many years of practical experience and comparison with reference cases have shown that the code together with the built-in data libraries gives reliable results. OMEGA is completely independent on other widely used criticality codes (KENO, MCNP, etc.), concerning programming and the data base. It is a good practice to run difficult criticality safety problems by different independent codes in order to mutually verify the results. In this way, OMEGA can be used as a redundant code within the family of criticality codes. An advantage of OMEGA is the short calculation time: A typical criticality safety application takes only a few minutes on a Pentium PC. Therefore, the influence of parameter variations can simply be investigated by running many variants of a problem. (orig.)
MCB. A continuous energy Monte Carlo burnup simulation code
International Nuclear Information System (INIS)
Cetnar, J.; Wallenius, J.; Gudowski, W.
1999-01-01
A code for integrated simulation of neutrinos and burnup based upon continuous energy Monte Carlo techniques and transmutation trajectory analysis has been developed. Being especially well suited for studies of nuclear waste transmutation systems, the code is an extension of the well validated MCNP transport program of Los Alamos National Laboratory. Among the advantages of the code (named MCB) is a fully integrated data treatment combined with a time-stepping routine that automatically corrects for burnup dependent changes in reaction rates, neutron multiplication, material composition and self-shielding. Fission product yields are treated as continuous functions of incident neutron energy, using a non-equilibrium thermodynamical model of the fission process. In the present paper a brief description of the code and applied methods are given. (author)
Energy Technology Data Exchange (ETDEWEB)
Vilches, M. [Servicio de Fisica y Proteccion Radiologica, Hospital Regional Universitario ' Virgen de las Nieves' , Avda. de las Fuerzas Armadas, 2, E-18014 Granada (Spain)], E-mail: mvilches@ugr.es; Garcia-Pareja, S. [Servicio de Radiofisica Hospitalaria, Hospital Regional Universitario ' Carlos Haya' , Avda. Carlos Haya, s/n, E-29010 Malaga (Spain); Guerrero, R. [Servicio de Radiofisica, Hospital Universitario ' San Cecilio' , Avda. Dr. Oloriz, 16, E-18012 Granada (Spain); Anguiano, M.; Lallena, A.M. [Departamento de Fisica Atomica, Molecular y Nuclear, Universidad de Granada, E-18071 Granada (Spain)
2007-09-21
When a therapeutic electron linear accelerator is simulated using a Monte Carlo (MC) code, the tuning of the initial spectra and the renormalization of dose (e.g., to maximum axial dose) constitute a common practice. As a result, very similar depth dose curves are obtained for different MC codes. However, if renormalization is turned off, the results obtained with the various codes disagree noticeably. The aim of this work is to investigate in detail the reasons of this disagreement. We have found that the observed differences are due to non-negligible differences in the angular scattering of the electron beam in very thin slabs of dense material (primary foil) and thick slabs of very low density material (air). To gain insight, the effects of the angular scattering models considered in various MC codes on the dose distribution in a water phantom are discussed using very simple geometrical configurations for the LINAC. The MC codes PENELOPE 2003, PENELOPE 2005, GEANT4, GEANT3, EGSnrc and MCNPX have been used.
International Nuclear Information System (INIS)
Vilches, M.; Garcia-Pareja, S.; Guerrero, R.; Anguiano, M.; Lallena, A.M.
2007-01-01
When a therapeutic electron linear accelerator is simulated using a Monte Carlo (MC) code, the tuning of the initial spectra and the renormalization of dose (e.g., to maximum axial dose) constitute a common practice. As a result, very similar depth dose curves are obtained for different MC codes. However, if renormalization is turned off, the results obtained with the various codes disagree noticeably. The aim of this work is to investigate in detail the reasons of this disagreement. We have found that the observed differences are due to non-negligible differences in the angular scattering of the electron beam in very thin slabs of dense material (primary foil) and thick slabs of very low density material (air). To gain insight, the effects of the angular scattering models considered in various MC codes on the dose distribution in a water phantom are discussed using very simple geometrical configurations for the LINAC. The MC codes PENELOPE 2003, PENELOPE 2005, GEANT4, GEANT3, EGSnrc and MCNPX have been used
Proton therapy Monte Carlo SRNA-VOX code
Directory of Open Access Journals (Sweden)
Ilić Radovan D.
2012-01-01
Full Text Available The most powerful feature of the Monte Carlo method is the possibility of simulating all individual particle interactions in three dimensions and performing numerical experiments with a preset error. These facts were the motivation behind the development of a general-purpose Monte Carlo SRNA program for proton transport simulation in technical systems described by standard geometrical forms (plane, sphere, cone, cylinder, cube. Some of the possible applications of the SRNA program are: (a a general code for proton transport modeling, (b design of accelerator-driven systems, (c simulation of proton scattering and degrading shapes and composition, (d research on proton detectors; and (e radiation protection at accelerator installations. This wide range of possible applications of the program demands the development of various versions of SRNA-VOX codes for proton transport modeling in voxelized geometries and has, finally, resulted in the ISTAR package for the calculation of deposited energy distribution in patients on the basis of CT data in radiotherapy. All of the said codes are capable of using 3-D proton sources with an arbitrary energy spectrum in an interval of 100 keV to 250 MeV.
International Nuclear Information System (INIS)
Fernandez-Varea, J.M.; Salvat, F.; Liljequist, D.
1994-09-01
The details of a Monte Carlo code for computing the penetration and energy loss of electrons and positrons in solids are described. The code, intended for electrons and positrons with energies from ∼ 100 eV to ∼ 100 keV, is based on the simulation of individual elastic and inelastic collisions. Elastic collisions are simulated using differential cross sections computed by the relativistic partial wave method applied to a muffin-tin Dirac-Hartree-Fock-Slater potential. Inelastic collisions are simulated by means of a model based on optical and photoelectric data, which are extended to the non-zero momentum transfer region by means of somewhat different algorithms for valence electron excitations and inner-shell excitations. This report focuses on the description of detailed formulae and sampling methods. 10 refs, 3 figs, 8 tabs
Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN
International Nuclear Information System (INIS)
Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi; Kaneko, Toshiyuki.
1996-12-01
The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code 'MULTI-KENO' and the routine for the burnup calculation of the one dimensional burnup code 'UNITBURN'. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)
Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN
Energy Technology Data Exchange (ETDEWEB)
Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Toshiyuki
1996-12-01
The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code `MULTI-KENO` and the routine for the burnup calculation of the one dimensional burnup code `UNITBURN`. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)
Energy Technology Data Exchange (ETDEWEB)
Cupini, E. [ENEA, Centro Ricerche `Ezio Clementel`, Bologna (Italy). Dipt. Innovazione; Borgia, M.G. [ENEA, Centro Ricerche `Ezio Clementel`, Bologna (Italy). Dipt. Energia; Premuda, M. [Consiglio Nazionale delle Ricerche, Bologna (Italy). Ist. FISBAT
1997-03-01
The Montecarlo code PREMAR is described, which allows the user to simulate the radiation transport in the atmosphere, in the ultraviolet-infrared frequency interval. A plan multilayer geometry is at present foreseen by the code, witch albedo possibility at the lower boundary surface. For a given monochromatic point source, the main quantities computed by the code are the absorption spatial distributions of aerosol and molecules, together with the related atmospheric transmittances. Moreover, simulation of of Lidar experiments are foreseen by the code, the source and telescope fields of view being assigned. To build-up the appropriate probability distributions, an input data library is assumed to be read by the code. For this purpose the radiance-transmittance LOWTRAN-7 code has been conveniently adapted as a source of the library so as to exploit the richness of information of the code for a large variety of atmospheric simulations. Results of applications of the PREMAR code are finally presented, with special reference to simulations of Lidar system and radiometer experiments carried out at the Brasimone ENEA Centre by the Environment Department.
Nuclear reactions in Monte Carlo codes
Ferrari, Alfredo
2002-01-01
The physics foundations of hadronic interactions as implemented in most Monte Carlo codes are presented together with a few practical examples. The description of the relevant physics is presented schematically split into the major steps in order to stress the different approaches required for the full understanding of nuclear reactions at intermediate and high energies. Due to the complexity of the problem, only a few semi-qualitative arguments are developed in this paper. The description will be necessarily schematic and somewhat incomplete, but hopefully it will be useful for a first introduction into this topic. Examples are shown mostly for the high energy regime, where all mechanisms mentioned in the paper are at work and to which perhaps most of the readers are less accustomed. Examples for lower energies can be found in the references. (43 refs) .
Applications guide to the MORSE Monte Carlo code
International Nuclear Information System (INIS)
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
Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method
2002-01-01
This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.
Feasibility Study of Core Design with a Monte Carlo Code for APR1400 Initial core
Energy Technology Data Exchange (ETDEWEB)
Kim, Jinsun; Chang, Do Ik; Seong, Kibong [KEPCO NF, Daejeon (Korea, Republic of)
2014-10-15
The Monte Carlo calculation becomes more popular and useful nowadays due to the rapid progress in computing power and parallel calculation techniques. There have been many attempts to analyze a commercial core by Monte Carlo transport code using the enhanced computer capability, recently. In this paper, Monte Carlo calculation of APR1400 initial core has been performed and the results are compared with the calculation results of conventional deterministic code to find out the feasibility of core design using Monte Carlo code. SERPENT, a 3D continuous-energy Monte Carlo reactor physics burnup calculation code is used for this purpose and the KARMA-ASTRA code system, which is used for a deterministic code of comparison. The preliminary investigation for the feasibility of commercial core design with Monte Carlo code was performed in this study. Simplified core geometry modeling was performed for the reactor core surroundings and reactor coolant model is based on two region model. The reactivity difference at HZP ARO condition between Monte Carlo code and the deterministic code is consistent with each other and the reactivity difference during the depletion could be reduced by adopting the realistic moderator temperature. The reactivity difference calculated at HFP, BOC, ARO equilibrium condition was 180 ±9 pcm, with axial moderator temperature of a deterministic code. The computing time will be a significant burden at this time for the application of Monte Carlo code to the commercial core design even with the application of parallel computing because numerous core simulations are required for actual loading pattern search. One of the remedy will be a combination of Monte Carlo code and the deterministic code to generate the physics data. The comparison of physics parameters with sophisticated moderator temperature modeling and depletion will be performed for a further study.
Modifications to the Monte Carlo neutronics code MONK
International Nuclear Information System (INIS)
Hutton, J.L.
1979-09-01
The Monte Carlo neutronics code MONK has been widely used for criticality calculations, and is one of the standard methods for assessing the safety of transport flasks and fuel storage facilities in the UK. Recently, attempts have been made to extend the range of applications of this calculational technique. In particular studies have been carried out using Monte Carlo to analyse reactor physics experiments. In these applications various shortcomings of the standard version MONK5 became apparent. The basic data library was found to be inadequate and additional estimates of parameters (eg power distribution) not normally included in criticality studies were required. These features which required improvement, primarily in the context of using the code for reactor physics calculations, are enumerated. To facilitate the use of the code as a reactor physics calculational tool a series of modifications have been carried out. The code has been modified so that the user can use group data tabulations of the cross sections instead of the present 'point' data values. The code can now interface with a number of reactor physics group data preparation schemes but in particular it can use WIMS-E interfaces as a source of group data. Details of the changes are outlined and a new version of MONK incorporating these modifications has been created. This version is called MONK5W. This paper provides a guide to the use of this version. The data input is described along with other details required to use this code on the Harwell IBM 3033. To aid the user, examples of calculations using the new facilities incorporated in MONK5W are given. (UK)
Energy Technology Data Exchange (ETDEWEB)
NONE
2001-01-01
In the report, research results discussed in 1999 fiscal year at Nuclear Code Evaluation Committee of Nuclear Code Research Committee were summarized. Present status of Monte Carlo simulation on nuclear energy study was described. Especially, besides of criticality, shielding and core analyses, present status of applications to risk and radiation damage analyses, high energy transport and nuclear theory calculations of Monte Carlo Method was described. The 18 papers are indexed individually. (J.P.N.)
Recent developments in the Los Alamos radiation transport code system
International Nuclear Information System (INIS)
Forster, R.A.; Parsons, K.
1997-01-01
A brief progress report on updates to the Los Alamos Radiation Transport Code System (LARTCS) for solving criticality and fixed-source problems is provided. LARTCS integrates the Diffusion Accelerated Neutral Transport (DANT) discrete ordinates codes with the Monte Carlo N-Particle (MCNP) code. The LARCTS code is being developed with a graphical user interface for problem setup and analysis. Progress in the DANT system for criticality applications include a two-dimensional module which can be linked to a mesh-generation code and a faster iteration scheme. Updates to MCNP Version 4A allow statistical checks of calculated Monte Carlo results
Parallel MCNP Monte Carlo transport calculations with MPI
International Nuclear Information System (INIS)
Wagner, J.C.; Haghighat, A.
1996-01-01
The steady increase in computational performance has made Monte Carlo calculations for large/complex systems possible. However, in order to make these calculations practical, order of magnitude increases in performance are necessary. The Monte Carlo method is inherently parallel (particles are simulated independently) and thus has the potential for near-linear speedup with respect to the number of processors. Further, the ever-increasing accessibility of parallel computers, such as workstation clusters, facilitates the practical use of parallel Monte Carlo. Recognizing the nature of the Monte Carlo method and the trends in available computing, the code developers at Los Alamos National Laboratory implemented the message-passing general-purpose Monte Carlo radiation transport code MCNP (version 4A). The PVM package was chosen by the MCNP code developers because it supports a variety of communication networks, several UNIX platforms, and heterogeneous computer systems. This PVM version of MCNP has been shown to produce speedups that approach the number of processors and thus, is a very useful tool for transport analysis. Due to software incompatibilities on the local IBM SP2, PVM has not been available, and thus it is not possible to take advantage of this useful tool. Hence, it became necessary to implement an alternative message-passing library package into MCNP. Because the message-passing interface (MPI) is supported on the local system, takes advantage of the high-speed communication switches in the SP2, and is considered to be the emerging standard, it was selected
Weighted-delta-tracking for Monte Carlo particle transport
International Nuclear Information System (INIS)
Morgan, L.W.G.; Kotlyar, D.
2015-01-01
Highlights: • This paper presents an alteration to the Monte Carlo Woodcock tracking technique. • The alteration improves computational efficiency within regions of high absorbers. • The rejection technique is replaced by a statistical weighting mechanism. • The modified Woodcock method is shown to be faster than standard Woodcock tracking. • The modified Woodcock method achieves a lower variance, given a specified accuracy. - Abstract: Monte Carlo particle transport (MCPT) codes are incredibly powerful and versatile tools to simulate particle behavior in a multitude of scenarios, such as core/criticality studies, radiation protection, shielding, medicine and fusion research to name just a small subset applications. However, MCPT codes can be very computationally expensive to run when the model geometry contains large attenuation depths and/or contains many components. This paper proposes a simple modification to the Woodcock tracking method used by some Monte Carlo particle transport codes. The Woodcock method utilizes the rejection method for sampling virtual collisions as a method to remove collision distance sampling at material boundaries. However, it suffers from poor computational efficiency when the sample acceptance rate is low. The proposed method removes rejection sampling from the Woodcock method in favor of a statistical weighting scheme, which improves the computational efficiency of a Monte Carlo particle tracking code. It is shown that the modified Woodcock method is less computationally expensive than standard ray-tracing and rejection-based Woodcock tracking methods and achieves a lower variance, given a specified accuracy
High Energy Transport Code HETC
International Nuclear Information System (INIS)
Gabriel, T.A.
1985-09-01
The physics contained in the High Energy Transport Code (HETC), in particular the collision models, are discussed. An application using HETC as part of the CALOR code system is also given. 19 refs., 5 figs., 3 tabs
DEFF Research Database (Denmark)
Hansen, Jonas; Krigslund, Jeppe; Roetter, Daniel Enrique Lucani
2014-01-01
Packet losses in wireless networks dramatically curbs the performance of TCP. This paper introduces a simple coding shim that aids IP-layer traffic in lossy environments while being transparent to transport layer protocols. The proposed coding approach enables erasure correction while being...... oblivious to the congestion control algorithms of the utilised transport layer protocol. Although our coding shim is indifferent towards the transport layer protocol, we focus on the performance of TCP when ran on top of our proposed coding mechanism due to its widespread use. The coding shim provides gains...
Baräo, Fernando; Nakagawa, Masayuki; Távora, Luis; Vaz, Pedro
2001-01-01
This book focusses on the state of the art of Monte Carlo methods in radiation physics and particle transport simulation and applications, the latter involving in particular, the use and development of electron--gamma, neutron--gamma and hadronic codes. Besides the basic theory and the methods employed, special attention is paid to algorithm development for modeling, and the analysis of experiments and measurements in a variety of fields ranging from particle to medical physics.
Development of Monte Carlo-based pebble bed reactor fuel management code
International Nuclear Information System (INIS)
Setiadipura, Topan; Obara, Toru
2014-01-01
Highlights: • A new Monte Carlo-based fuel management code for OTTO cycle pebble bed reactor was developed. • The double-heterogeneity was modeled using statistical method in MVP-BURN code. • The code can perform analysis of equilibrium and non-equilibrium phase. • Code-to-code comparisons for Once-Through-Then-Out case were investigated. • Ability of the code to accommodate the void cavity was confirmed. - Abstract: A fuel management code for pebble bed reactors (PBRs) based on the Monte Carlo method has been developed in this study. The code, named Monte Carlo burnup analysis code for PBR (MCPBR), enables a simulation of the Once-Through-Then-Out (OTTO) cycle of a PBR from the running-in phase to the equilibrium condition. In MCPBR, a burnup calculation based on a continuous-energy Monte Carlo code, MVP-BURN, is coupled with an additional utility code to be able to simulate the OTTO cycle of PBR. MCPBR has several advantages in modeling PBRs, namely its Monte Carlo neutron transport modeling, its capability of explicitly modeling the double heterogeneity of the PBR core, and its ability to model different axial fuel speeds in the PBR core. Analysis at the equilibrium condition of the simplified PBR was used as the validation test of MCPBR. The calculation results of the code were compared with the results of diffusion-based fuel management PBR codes, namely the VSOP and PEBBED codes. Using JENDL-4.0 nuclide library, MCPBR gave a 4.15% and 3.32% lower k eff value compared to VSOP and PEBBED, respectively. While using JENDL-3.3, MCPBR gave a 2.22% and 3.11% higher k eff value compared to VSOP and PEBBED, respectively. The ability of MCPBR to analyze neutron transport in the top void of the PBR core and its effects was also confirmed
Simulation of transport equations with Monte Carlo
International Nuclear Information System (INIS)
Matthes, W.
1975-09-01
The main purpose of the report is to explain the relation between the transport equation and the Monte Carlo game used for its solution. The introduction of artificial particles carrying a weight provides one with high flexibility in constructing many different games for the solution of the same equation. This flexibility opens a way to construct a Monte Carlo game for the solution of the adjoint transport equation. Emphasis is laid mostly on giving a clear understanding of what to do and not on the details of how to do a specific game
DOSE COEFFICIENTS FOR LIVER CHEMOEMBOLISATION PROCEDURES USING MONTE CARLO CODE.
Karavasilis, E; Dimitriadis, A; Gonis, H; Pappas, P; Georgiou, E; Yakoumakis, E
2016-12-01
The aim of the present study is the estimation of radiation burden during liver chemoembolisation procedures. Organ dose and effective dose conversion factors, normalised to dose-area product (DAP), were estimated for chemoembolisation procedures using a Monte Carlo transport code in conjunction with an adult mathematical phantom. Exposure data from 32 patients were used to determine the exposure projections for the simulations. Equivalent organ (H T ) and effective (E) doses were estimated using individual DAP values. The organs receiving the highest amount of doses during these exams were lumbar spine, liver and kidneys. The mean effective dose conversion factor was 1.4 Sv Gy -1 m -2 Dose conversion factors can be useful for patient-specific radiation burden during chemoembolisation procedures. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.
Monte Carlo code criticality benchmark comparisons for waste packaging
International Nuclear Information System (INIS)
Alesso, H.P.; Annese, C.E.; Buck, R.M.; Pearson, J.S.; Lloyd, W.R.
1992-07-01
COG is a new point-wise Monte Carlo code being developed and tested at Lawrence Livermore National Laboratory (LLNL). It solves the Boltzmann equation for the transport of neutrons and photons. The objective of this paper is to report on COG results for criticality benchmark experiments both on a Cray mainframe and on a HP 9000 workstation. COG has been recently ported to workstations to improve its accessibility to a wider community of users. COG has some similarities to a number of other computer codes used in the shielding and criticality community. The recently introduced high performance reduced instruction set (RISC) UNIX workstations provide computational power that approach mainframes at a fraction of the cost. A version of COG is currently being developed for the Hewlett Packard 9000/730 computer with a UNIX operating system. Subsequent porting operations will move COG to SUN, DEC, and IBM workstations. In addition, a CAD system for preparation of the geometry input for COG is being developed. In July 1977, Babcock ampersand Wilcox Co. (B ampersand W) was awarded a contract to conduct a series of critical experiments that simulated close-packed storage of LWR-type fuel. These experiments provided data for benchmarking and validating calculational methods used in predicting K-effective of nuclear fuel storage in close-packed, neutron poisoned arrays. Low enriched UO2 fuel pins in water-moderated lattices in fuel storage represent a challenging criticality calculation for Monte Carlo codes particularly when the fuel pins extend out of the water. COG and KENO calculational results of these criticality benchmark experiments are presented
Fast code for Monte Carlo simulations
International Nuclear Information System (INIS)
Oliveira, P.M.C. de; Penna, T.J.P.
1988-01-01
A computer code to generate the dynamic evolution of the Ising model on a square lattice, following the Metropolis algorithm is presented. The computer time consumption is reduced by a factor of 8 when one compares our code with traditional multiple spin codes. The memory allocation size is also reduced by a factor of 4. The code is easily generalizable for other lattices and models. (author) [pt
Scalable Domain Decomposed Monte Carlo Particle Transport
Energy Technology Data Exchange (ETDEWEB)
O' Brien, Matthew Joseph [Univ. of California, Davis, CA (United States)
2013-12-05
In this dissertation, we present the parallel algorithms necessary to run domain decomposed Monte Carlo particle transport on large numbers of processors (millions of processors). Previous algorithms were not scalable, and the parallel overhead became more computationally costly than the numerical simulation.
International Nuclear Information System (INIS)
Martin, William R.; Brown, Forrest B.
2001-01-01
We present an alternative Monte Carlo method for solving the coupled equations of radiation transport and material energy. This method is based on incorporating the analytical solution to the material energy equation directly into the Monte Carlo simulation for the radiation intensity. This method, which we call the Analytical Monte Carlo (AMC) method, differs from the well known Implicit Monte Carlo (IMC) method of Fleck and Cummings because there is no discretization of the material energy equation since it is solved as a by-product of the Monte Carlo simulation of the transport equation. Our method also differs from the method recently proposed by Ahrens and Larsen since they use Monte Carlo to solve both equations, while we are solving only the radiation transport equation with Monte Carlo, albeit with effective sources and cross sections to represent the emission sources. Our method bears some similarity to a method developed and implemented by Carter and Forest nearly three decades ago, but there are substantive differences. We have implemented our method in a simple zero-dimensional Monte Carlo code to test the feasibility of the method, and the preliminary results are very promising, justifying further extension to more realistic geometries. (authors)
Gamma streaming experiments for validation of Monte Carlo code
International Nuclear Information System (INIS)
Thilagam, L.; Mohapatra, D.K.; Subbaiah, K.V.; Iliyas Lone, M.; Balasubramaniyan, V.
2012-01-01
In-homogeneities in shield structures lead to considerable amount of leakage radiation (streaming) increasing the radiation levels in accessible areas. Development works on experimental as well as computational methods for quantifying this streaming radiation are still continuing. Monte Carlo based radiation transport code, MCNP is usually a tool for modeling and analyzing such problems involving complex geometries. In order to validate this computational method for streaming analysis, it is necessary to carry out some experimental measurements simulating these inhomogeneities like ducts and voids present in the bulk shields for typical cases. The data thus generated will be analysed by simulating the experimental set up employing MCNP code and optimized input parameters for the code in finding solutions for similar radiation streaming problems will be formulated. Comparison of experimental data obtained from radiation streaming experiments through ducts will give a set of thumb rules and analytical fits for total radiation dose rates within and outside the duct. The present study highlights the validation of MCNP code through the gamma streaming experiments carried out with the ducts of various shapes and dimensions. Over all, the present study throws light on suitability of MCNP code for the analysis of gamma radiation streaming problems for all duct configurations considered. In the present study, only dose rate comparisons have been made. Studies on spectral comparison of streaming radiation are in process. Also, it is planned to repeat the experiments with various shield materials. Since the penetrations and ducts through bulk shields are unavoidable in an operating nuclear facility the results on this kind of radiation streaming simulations and experiments will be very useful in the shield structure optimization without compromising the radiation safety
International Nuclear Information System (INIS)
Gosmain, Cecile-Aline
2011-01-01
In the framework of French research program on Generation IV sodium cooled fast reactor, one possible option consists in burning minor actinides in this kind of Advanced Sodium Technological Reactor. Two types of transmutation mode are studied in the world : the homogeneous mode of transmutation where actinides are scattered with very low enrichment ratio in fissile assemblies and the heterogeneous mode where fissile core is surrounded by blanket assemblies filled with minor actinides with ratio of incorporated actinides up to 20%. Depending on which element is considered to be burnt and on its content, these minor actinides contents imply constraints on assemblies' transportation between Nuclear Power Plants and fuel cycle facilities. In this study, we present some academic studies in order to identify some key constraints linked to the residual power and neutron/gamma load of such kind of blanket assemblies. To simplify the approach, we considered a modeling of a 'model cask' dedicated to the transportation of a unique irradiated blanket assembly loaded with 20% of Americium and basically inspired from an existent cask designed initially for the damaged fissile Superphenix assembly transport. Thermal calculations performed with EDF-SYRTHES code have shown that due to thermal limitations on cladding temperature, the decay time to be considered before transportation is 20 years. This study is based on explicit 3D representations of the cask and the contained blanket assembly with the Monte Carlo code TRIPOLI/JEFF3.1.1 library and concludes that after such a decay time, the transportation of a unique Americium radial blanket is feasible only if the design of our model cask is modified in order to comply with the dose limitation criterion. (author)
Comparison of calculations of a reflected reactor with diffusion, SN and Monte Carlo codes
International Nuclear Information System (INIS)
McGregor, B.
1975-01-01
A diffusion theory code, POW, was compared with a Monte Carlo transport theory code, KENO, for the calculation of a small C/ 235 U cylindrical core with a graphite reflector. The calculated multiplication factors were in good agreement but differences were noted in region-averaged group fluxes. A one-dimensional spherical geometry was devised to approximate cylindrical geometry. Differences similar to those already observed were noted when the region-averaged fluxes from a diffusion theory (POW) calculation were compared with an SN transport theory (ANAUSN) calculation for the spherical model. Calculations made with SN and Monte Carlo transport codes were in good agreement. It was concluded that observed flux differences were attributable to the POW code, and were not inconsistent with inherent diffusion theory approximations. (author)
Monte Carlo radiation transport: A revolution in science
International Nuclear Information System (INIS)
Hendricks, J.
1993-01-01
When Enrico Fermi, Stan Ulam, Nicholas Metropolis, John von Neuman, and Robert Richtmyer invented the Monte Carlo method fifty years ago, little could they imagine the far-flung consequences, the international applications, and the revolution in science epitomized by their abstract mathematical method. The Monte Carlo method is used in a wide variety of fields to solve exact computational models approximately by statistical sampling. It is an alternative to traditional physics modeling methods which solve approximate computational models exactly by deterministic methods. Modern computers and improved methods, such as variance reduction, have enhanced the method to the point of enabling a true predictive capability in areas such as radiation or particle transport. This predictive capability has contributed to a radical change in the way science is done: design and understanding come from computations built upon experiments rather than being limited to experiments, and the computer codes doing the computations have become the repository for physics knowledge. The MCNP Monte Carlo computer code effort at Los Alamos is an example of this revolution. Physicians unfamiliar with physics details can design cancer treatments using physics buried in the MCNP computer code. Hazardous environments and hypothetical accidents can be explored. Many other fields, from underground oil well exploration to aerospace, from physics research to energy production, from safety to bulk materials processing, benefit from MCNP, the Monte Carlo method, and the revolution in science
Monte Carlo simulation of medical linear accelerator using primo code
International Nuclear Information System (INIS)
Omer, Mohamed Osman Mohamed Elhasan
2014-12-01
The use of monte Carlo simulation has become very important in the medical field and especially in calculation in radiotherapy. Various Monte Carlo codes were developed simulating interactions of particles and photons with matter. One of these codes is PRIMO that performs simulation of radiation transport from the primary electron source of a linac to estimate the absorbed dose in a water phantom or computerized tomography (CT). PRIMO is based on Penelope Monte Carlo code. Measurements of 6 MV photon beam PDD and profile were done for Elekta precise linear accelerator at Radiation and Isotopes Center Khartoum using computerized Blue water phantom and CC13 Ionization Chamber. accept Software was used to control the phantom to measure and verify dose distribution. Elektalinac from the list of available linacs in PRIMO was tuned to model Elekta precise linear accelerator. Beam parameter of 6.0 MeV initial electron energy, 0.20 MeV FWHM, and 0.20 cm focal spot FWHM were used, and an error of 4% between calculated and measured curves was found. The buildup region Z max was 1.40 cm and homogenous profile in cross line and in line were acquired. A number of studies were done to verily the model usability one of them is the effect of the number of histories on accuracy of the simulation and the resulted profile for the same beam parameters. The effect was noticeable and inaccuracies in the profile were reduced by increasing the number of histories. Another study was the effect of Side-step errors on the calculated dose which was compared with the measured dose for the same setting.It was in range of 2% for 5 cm shift, but it was higher in the calculated dose because of the small difference between the tuned model and measured dose curves. Future developments include simulating asymmetrical fields, calculating the dose distribution in computerized tomographic (CT) volume, studying the effect of beam modifiers on beam profile for both electron and photon beams.(Author)
MORET: Version 4.B. A multigroup Monte Carlo criticality code
International Nuclear Information System (INIS)
Jacquet, Olivier; Miss, Joachim; Courtois, Gerard
2003-01-01
MORET 4 is a three dimensional multigroup Monte Carlo code which calculates the effective multiplication factor (keff) of any configurations more or less complex as well as reaction rates in the different volumes of the geometry and the leakage out of the system. MORET 4 is the Monte Carlo code of the APOLLO2-MORET 4 standard route of CRISTAL, the French criticality package. It is the most commonly used Monte Carlo code for French criticality calculations. During the last four years, the MORET 4 team has developed or improved the following major points: modernization of the geometry, implementation of perturbation algorithms, source distribution convergence, statistical detection of stationarity, unbiased variance estimation and creation of pre-processing and post-processing tools. The purpose of this paper is not only to present the new features of MORET but also to detail clearly the physical models and the mathematical methods used in the code. (author)
Application of the three-dimensional Oak Ridge transport code
International Nuclear Information System (INIS)
Rhoades, W.A.; Childs, R.L.; Emmett, M.B.; Cramer, S.N.
1984-01-01
TORT, a 3-d extension of the DOT discrete ordinates transport code is now in production use for studies of radiation penetration into large concrete and masonry structures. This paper discusses certain features of the new code and shows representative results, including comparisons with Monte Carlo calculations
Applications of the Los Alamos High Energy Transport code
International Nuclear Information System (INIS)
Waters, L.; Gavron, A.; Prael, R.E.
1992-01-01
Simulation codes reliable through a large range of energies are essential to analyze the environment of vehicles and habitats proposed for space exploration. The LAHET monte carlo code has recently been expanded to track high energy hadrons with FLUKA, while retaining the original Los Alamos version of HETC at lower energies. Electrons and photons are transported with EGS4, and an interface to the MCNP monte carlo code is provided to analyze neutrons with kinetic energies less than 20 MeV. These codes are benchmarked by comparison of LAHET/MCNP calculations to data from the Brookhaven experiment E814 participant calorimeter
Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations
International Nuclear Information System (INIS)
Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I.
2015-01-01
A burn-up calculation of VVER's cores by Monte-Carlo code is complex process and requires large computational costs. This fact makes Monte-Carlo codes usage complicated for project and operating calculations. Previously prepared isotopic compositions are proposed to use for the Monte-Carlo code (MCU) calculations of different states of VVER's core with burnt fuel. Isotopic compositions are proposed to calculate by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by engineering codes (TVS-M, PERMAK-A). The multiplication factors and power distributions of FA and VVER with infinite height are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The MCU calculation data were compared with the data which were obtained by engineering codes.
Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System
Karim, Julia Abdul
2008-05-01
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.
Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System
International Nuclear Information System (INIS)
Karim, Julia Abdul
2008-01-01
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
Summary - COG: A new point-wise Monte Carlo code for burnup credit analysis
International Nuclear Information System (INIS)
Alesso, H.P.
1989-01-01
COG, a new point-wise Monte Carlo code being developed and tested at Lawrence Livermore National Laboratory (LLNL) for the Cray-1, solves the Boltzmann equation for the transport of neutrons, photons, and (in future versions) other particles. Techniques included in the code for modifying the random walk of particles make COG most suitable for solving deep-penetration (shielding) problems and a wide variety of criticality problems. COG is similar to a number of other computer codes used in the shielding community. Each code is a little different in its geometry input and its random-walk modification options. COG is a Monte Carlo code specifically designed for the CRAY (in 1986) to be as precise as the current state of physics knowledge. It has been extensively benchmarked and used as a shielding code at LLNL since 1986, and has recently been extended to accomplish criticality calculations. It will make an excellent tool for future shipping cask studies
A Monte Carlo track structure code for low energy protons
Endo, S; Nikjoo, H; Uehara, S; Hoshi, M; Ishikawa, M; Shizuma, K
2002-01-01
A code is described for simulation of protons (100 eV to 10 MeV) track structure in water vapor. The code simulates molecular interaction by interaction for the transport of primary ions and secondary electrons in the form of ionizations and excitations. When a low velocity ion collides with the atoms or molecules of a target, the ion may also capture or lose electrons. The probabilities for these processes are described by the quantity cross-section. Although proton track simulation at energies above Bragg peak (>0.3 MeV) has been achieved to a high degree of precision, simulations at energies near or below the Bragg peak have only been attempted recently because of the lack of relevant cross-section data. As the hydrogen atom has a different ionization cross-section from that of a proton, charge exchange processes need to be considered in order to calculate stopping power for low energy protons. In this paper, we have used state-of-the-art Monte Carlo track simulation techniques, in conjunction with the pub...
Los Alamos radiation transport code system on desktop computing platforms
International Nuclear Information System (INIS)
Briesmeister, J.F.; Brinkley, F.W.; Clark, B.A.; West, J.T.
1990-01-01
The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. These codes were originally developed many years ago and have undergone continual improvement. With a large initial effort and continued vigilance, the codes are easily portable from one type of hardware to another. The performance of scientific work-stations (SWS) has evolved to the point that such platforms can be used routinely to perform sophisticated radiation transport calculations. As the personal computer (PC) performance approaches that of the SWS, the hardware options for desk-top radiation transport calculations expands considerably. The current status of the radiation transport codes within the LARTCS is described: MCNP, SABRINA, LAHET, ONEDANT, TWODANT, TWOHEX, and ONELD. Specifically, the authors discuss hardware systems on which the codes run and present code performance comparisons for various machines
Cost of splitting in Monte Carlo transport
International Nuclear Information System (INIS)
Everett, C.J.; Cashwell, E.D.
1978-03-01
In a simple transport problem designed to estimate transmission through a plane slab of x free paths by Monte Carlo methods, it is shown that m-splitting (m > or = 2) does not pay unless exp(x) > m(m + 3)/(m - 1). In such a case, the minimum total cost in terms of machine time is obtained as a function of m, and the optimal value of m is determined
MCOR - Monte Carlo depletion code for reference LWR calculations
Energy Technology Data Exchange (ETDEWEB)
Puente Espel, Federico, E-mail: fup104@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Tippayakul, Chanatip, E-mail: cut110@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Ivanov, Kostadin, E-mail: kni1@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Misu, Stefan, E-mail: Stefan.Misu@areva.com [AREVA, AREVA NP GmbH, Erlangen (Germany)
2011-04-15
Research highlights: > Introduction of a reference Monte Carlo based depletion code with extended capabilities. > Verification and validation results for MCOR. > Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations. Additionally
MCOR - Monte Carlo depletion code for reference LWR calculations
International Nuclear Information System (INIS)
Puente Espel, Federico; Tippayakul, Chanatip; Ivanov, Kostadin; Misu, Stefan
2011-01-01
Research highlights: → Introduction of a reference Monte Carlo based depletion code with extended capabilities. → Verification and validation results for MCOR. → Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations
International Nuclear Information System (INIS)
Turner, J.E.; Modolo, J.T.; Sordi, G.M.A.A.; Hamm, R.N.; Wright, H.A.
1979-01-01
PHOEL provides a source term for a Monte Carlo code which calculates the electron transport and energy degradation in liquid water. This code is used to study the relative biological effectiveness (RBE) of low-LET radiation at low doses. The basic numerical data used and their mathematical treatment are described as well as the operation of the code [pt
Monte Carlo method in radiation transport problems
International Nuclear Information System (INIS)
Dejonghe, G.; Nimal, J.C.; Vergnaud, T.
1986-11-01
In neutral radiation transport problems (neutrons, photons), two values are important: the flux in the phase space and the density of particles. To solve the problem with Monte Carlo method leads to, among other things, build a statistical process (called the play) and to provide a numerical value to a variable x (this attribution is called score). Sampling techniques are presented. Play biasing necessity is proved. A biased simulation is made. At last, the current developments (rewriting of programs for instance) are presented due to several reasons: two of them are the vectorial calculation apparition and the photon and neutron transport in vacancy media [fr
Energy Technology Data Exchange (ETDEWEB)
Yan, Qiang [College of Nuclear Science and Technology, Harbin Engineering University, Harbin 150001 (China); Shao, Lin, E-mail: lshao@tamu.edu [Department of Nuclear Engineering, Texas A& M University, College Station, TX 77843 (United States)
2017-03-15
Current popular Monte Carlo simulation codes for simulating electron bombardment in solids focus primarily on electron trajectories, instead of electron-induced displacements. Here we report a Monte Carol simulation code, DEEPER (damage creation and particle transport in matter), developed for calculating 3-D distributions of displacements produced by electrons of incident energies up to 900 MeV. Electron elastic scattering is calculated by using full-Mott cross sections for high accuracy, and primary-knock-on-atoms (PKAs)-induced damage cascades are modeled using ZBL potential. We compare and show large differences in 3-D distributions of displacements and electrons in electron-irradiated Fe. The distributions of total displacements are similar to that of PKAs at low electron energies. But they are substantially different for higher energy electrons due to the shifting of PKA energy spectra towards higher energies. The study is important to evaluate electron-induced radiation damage, for the applications using high flux electron beams to intentionally introduce defects and using an electron analysis beam for microstructural characterization of nuclear materials.
Verification of the shift Monte Carlo code with the C5G7 reactor benchmark
International Nuclear Information System (INIS)
Sly, N. C.; Mervin, B. T.; Mosher, S. W.; Evans, T. M.; Wagner, J. C.; Maldonado, G. I.
2012-01-01
Shift is a new hybrid Monte Carlo/deterministic radiation transport code being developed at Oak Ridge National Laboratory. At its current stage of development, Shift includes a parallel Monte Carlo capability for simulating eigenvalue and fixed-source multigroup transport problems. This paper focuses on recent efforts to verify Shift's Monte Carlo component using the two-dimensional and three-dimensional C5G7 NEA benchmark problems. Comparisons were made between the benchmark eigenvalues and those output by the Shift code. In addition, mesh-based scalar flux tally results generated by Shift were compared to those obtained using MCNP5 on an identical model and tally grid. The Shift-generated eigenvalues were within three standard deviations of the benchmark and MCNP5-1.60 values in all cases. The flux tallies generated by Shift were found to be in very good agreement with those from MCNP. (authors)
PEREGRINE: An all-particle Monte Carlo code for radiation therapy
International Nuclear Information System (INIS)
Hartmann Siantar, C.L.; Chandler, W.P.; Rathkopf, J.A.; Svatos, M.M.; White, R.M.
1994-09-01
The goal of radiation therapy is to deliver a lethal dose to the tumor while minimizing the dose to normal tissues. To carry out this task, it is critical to calculate correctly the distribution of dose delivered. Monte Carlo transport methods have the potential to provide more accurate prediction of dose distributions than currently-used methods. PEREGRINE is a new Monte Carlo transport code developed at Lawrence Livermore National Laboratory for the specific purpose of modeling the effects of radiation therapy. PEREGRINE transports neutrons, photons, electrons, positrons, and heavy charged-particles, including protons, deuterons, tritons, helium-3, and alpha particles. This paper describes the PEREGRINE transport code and some preliminary results for clinically relevant materials and radiation sources
A Monte Carlo code for ion beam therapy
Anaïs Schaeffer
2012-01-01
Initially developed for applications in detector and accelerator physics, the modern Fluka Monte Carlo code is now used in many different areas of nuclear science. Over the last 25 years, the code has evolved to include new features, such as ion beam simulations. Given the growing use of these beams in cancer treatment, Fluka simulations are being used to design treatment plans in several hadron-therapy centres in Europe. Fluka calculates the dose distribution for a patient treated at CNAO with proton beams. The colour-bar displays the normalized dose values. Fluka is a Monte Carlo code that very accurately simulates electromagnetic and nuclear interactions in matter. In the 1990s, in collaboration with NASA, the code was developed to predict potential radiation hazards received by space crews during possible future trips to Mars. Over the years, it has become the standard tool to investigate beam-machine interactions, radiation damage and radioprotection issues in the CERN accelerator com...
Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis
Hoogenboom, J. Eduard; Sjenitzer, Bart L.
2014-06-01
To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.
Application of OMEGA Monte Carlo codes for radiation therapy treatment planning
International Nuclear Information System (INIS)
Ayyangar, Komanduri M.; Jiang, Steve B.
1998-01-01
The accuracy of conventional dose algorithms for radiosurgery treatment planning is limited, due to the inadequate consideration of the lateral radiation transport and the difficulty of acquiring accurate dosimetric data for very small beams. In the present paper, some initial work on the application of Monte Carlo method in radiation treatment planning in general, and in radiosurgery treatment planning in particular, has been presented. Two OMEGA Monte Carlo codes, BEAM and DOSXYZ, are used. The BEAM code is used to simulate the transport of particles in the linac treatment head and radiosurgery collimator. A phase space file is obtained from the BEAM simulation for each collimator size. The DOSXYZ code is used to calculate the dose distribution in the patient's body reconstructed from CT slices using the phase space file as input. The accuracy of OMEGA Monte Carlo simulation for radiosurgery dose calculation is verified by comparing the calculated and measured basic dosimetric data for several radiosurgery beams and a 4 x 4 cm 2 conventional beam. The dose distributions for three clinical cases are calculated using OMEGA codes as the dose engine for an in-house developed radiosurgery treatment planning system. The verification using basic dosimetric data and the dose calculation for clinical cases demonstrate the feasibility of applying OMEGA Monte Carlo code system to radiosurgery treatment planning. (author)
A general purpose code for Monte Carlo simulations
International Nuclear Information System (INIS)
Wilcke, W.W.; Rochester Univ., NY
1984-01-01
A general-purpose computer code MONTHY has been written to perform Monte Carlo simulations of physical systems. To achieve a high degree of flexibility the code is organized like a general purpose computer, operating on a vector describing the time dependent state of the system under simulation. The instruction set of the 'computer' is defined by the user and is therefore adaptable to the particular problem studied. The organization of MONTHY allows iterative and conditional execution of operations. (orig.)
Monte Carlo methods in electron transport problems. Pt. 1
International Nuclear Information System (INIS)
Cleri, F.
1989-01-01
The condensed-history Monte Carlo method for charged particles transport is reviewed and discussed starting from a general form of the Boltzmann equation (Part I). The physics of the electronic interactions, together with some pedagogic example will be introduced in the part II. The lecture is directed to potential users of the method, for which it can be a useful introduction to the subject matter, and wants to establish the basis of the work on the computer code RECORD, which is at present in a developing stage
Monte Carlo simulation for the transport beamline
Energy Technology Data Exchange (ETDEWEB)
Romano, F.; Cuttone, G.; Jia, S. B.; Varisano, A. [INFN, Laboratori Nazionali del Sud, Via Santa Sofia 62, Catania (Italy); Attili, A.; Marchetto, F.; Russo, G. [INFN, Sezione di Torino, Via P.Giuria, 1 10125 Torino (Italy); Cirrone, G. A. P.; Schillaci, F.; Scuderi, V. [INFN, Laboratori Nazionali del Sud, Via Santa Sofia 62, Catania, Italy and Institute of Physics Czech Academy of Science, ELI-Beamlines project, Na Slovance 2, Prague (Czech Republic); Carpinelli, M. [INFN Sezione di Cagliari, c/o Dipartimento di Fisica, Università di Cagliari, Cagliari (Italy); Tramontana, A. [INFN, Laboratori Nazionali del Sud, Via Santa Sofia 62, Catania, Italy and Università di Catania, Dipartimento di Fisica e Astronomia, Via S. Sofia 64, Catania (Italy)
2013-07-26
In the framework of the ELIMED project, Monte Carlo (MC) simulations are widely used to study the physical transport of charged particles generated by laser-target interactions and to preliminarily evaluate fluence and dose distributions. An energy selection system and the experimental setup for the TARANIS laser facility in Belfast (UK) have been already simulated with the GEANT4 (GEometry ANd Tracking) MC toolkit. Preliminary results are reported here. Future developments are planned to implement a MC based 3D treatment planning in order to optimize shots number and dose delivery.
Monte Carlo simulation for the transport beamline
International Nuclear Information System (INIS)
Romano, F.; Cuttone, G.; Jia, S. B.; Varisano, A.; Attili, A.; Marchetto, F.; Russo, G.; Cirrone, G. A. P.; Schillaci, F.; Scuderi, V.; Carpinelli, M.; Tramontana, A.
2013-01-01
In the framework of the ELIMED project, Monte Carlo (MC) simulations are widely used to study the physical transport of charged particles generated by laser-target interactions and to preliminarily evaluate fluence and dose distributions. An energy selection system and the experimental setup for the TARANIS laser facility in Belfast (UK) have been already simulated with the GEANT4 (GEometry ANd Tracking) MC toolkit. Preliminary results are reported here. Future developments are planned to implement a MC based 3D treatment planning in order to optimize shots number and dose delivery
Computer codes in particle transport physics
International Nuclear Information System (INIS)
Pesic, M.
2004-01-01
Simulation of transport and interaction of various particles in complex media and wide energy range (from 1 MeV up to 1 TeV) is very complicated problem that requires valid model of a real process in nature and appropriate solving tool - computer code and data library. A brief overview of computer codes based on Monte Carlo techniques for simulation of transport and interaction of hadrons and ions in wide energy range in three dimensional (3D) geometry is shown. Firstly, a short attention is paid to underline the approach to the solution of the problem - process in nature - by selection of the appropriate 3D model and corresponding tools - computer codes and cross sections data libraries. Process of data collection and evaluation from experimental measurements and theoretical approach to establishing reliable libraries of evaluated cross sections data is Ion g, difficult and not straightforward activity. For this reason, world reference data centers and specialized ones are acknowledged, together with the currently available, state of art evaluated nuclear data libraries, as the ENDF/B-VI, JEF, JENDL, CENDL, BROND, etc. Codes for experimental and theoretical data evaluations (e.g., SAMMY and GNASH) together with the codes for data processing (e.g., NJOY, PREPRO and GRUCON) are briefly described. Examples of data evaluation and data processing to generate computer usable data libraries are shown. Among numerous and various computer codes developed in transport physics of particles, the most general ones are described only: MCNPX, FLUKA and SHIELD. A short overview of basic application of these codes, physical models implemented with their limitations, energy ranges of particles and types of interactions, is given. General information about the codes covers also programming language, operation system, calculation speed and the code availability. An example of increasing computation speed of running MCNPX code using a MPI cluster compared to the code sequential option
Study on random number generator in Monte Carlo code
International Nuclear Information System (INIS)
Oya, Kentaro; Kitada, Takanori; Tanaka, Shinichi
2011-01-01
The Monte Carlo code uses a sequence of pseudo-random numbers with a random number generator (RNG) to simulate particle histories. A pseudo-random number has its own period depending on its generation method and the period is desired to be long enough not to exceed the period during one Monte Carlo calculation to ensure the correctness especially for a standard deviation of results. The linear congruential generator (LCG) is widely used as Monte Carlo RNG and the period of LCG is not so long by considering the increasing rate of simulation histories in a Monte Carlo calculation according to the remarkable enhancement of computer performance. Recently, many kinds of RNG have been developed and some of their features are better than those of LCG. In this study, we investigate the appropriate RNG in a Monte Carlo code as an alternative to LCG especially for the case of enormous histories. It is found that xorshift has desirable features compared with LCG, and xorshift has a larger period, a comparable speed to generate random numbers, a better randomness, and good applicability to parallel calculation. (author)
Smans, Kristien; Zoetelief, Johannes; Verbrugge, Beatrijs; Haeck, Wim; Struelens, Lara; Vanhavere, Filip; Bosmans, Hilde
2010-05-01
The purpose of this study was to compare and validate three methods to simulate radiographic image detectors with the Monte Carlo software MCNP/MCNPX in a time efficient way. The first detector model was the standard semideterministic radiography tally, which has been used in previous image simulation studies. Next to the radiography tally two alternative stochastic detector models were developed: A perfect energy integrating detector and a detector based on the energy absorbed in the detector material. Validation of three image detector models was performed by comparing calculated scatter-to-primary ratios (SPRs) with the published and experimentally acquired SPR values. For mammographic applications, SPRs computed with the radiography tally were up to 44% larger than the published results, while the SPRs computed with the perfect energy integrating detectors and the blur-free absorbed energy detector model were, on the average, 0.3% (ranging from -3% to 3%) and 0.4% (ranging from -5% to 5%) lower, respectively. For general radiography applications, the radiography tally overestimated the measured SPR by as much as 46%. The SPRs calculated with the perfect energy integrating detectors were, on the average, 4.7% (ranging from -5.3% to -4%) lower than the measured SPRs, whereas for the blur-free absorbed energy detector model, the calculated SPRs were, on the average, 1.3% (ranging from -0.1% to 2.4%) larger than the measured SPRs. For mammographic applications, both the perfect energy integrating detector model and the blur-free energy absorbing detector model can be used to simulate image detectors, whereas for conventional x-ray imaging using higher energies, the blur-free energy absorbing detector model is the most appropriate image detector model. The radiography tally overestimates the scattered part and should therefore not be used to simulate radiographic image detectors.
International Nuclear Information System (INIS)
Copeland, K.; Parker, D. E.; Friedberg, W.
2010-01-01
Conversion coefficients were calculated for fluence-to-absorbed dose, fluence-to-equivalent dose, fluence-to-effective dose and fluence-to-gray equivalent for isotropic exposure of an adult female and an adult male to tritons ( 3 H + ) in the energy range of 10 MeV to 1 TeV (0.01-1000 GeV). Coefficients were calculated using Monte Carlo transport code MCNPX 2.7.C and BodyBuilder TM 1.3 anthropomorphic phantoms. Phantoms were modified to allow calculation of effective dose to a Reference Person using tissues and tissue weighting factors from 1990 and 2007 recommendations of the International Commission on Radiological Protection (ICRP) and calculation of gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. At 15 of the 19 energies for which coefficients for effective dose were calculated, coefficients based on ICRP 2007 and 1990 recommendations differed by less than 3%. The greatest difference, 43%, occurred at 30 MeV. Published by Oxford Univ. Press on behalf of the US Government 2010. (authors)
International Nuclear Information System (INIS)
Copeland, K.; Parker, D. E.; Friedberg, W.
2010-01-01
Conversion coefficients were calculated for fluence-to-absorbed dose, fluence-to-equivalent dose, fluence-to-effective dose and fluence-to-gray equivalent, for isotropic exposure of an adult male and an adult female to helions ( 3 He 2+ ) in the energy range of 10 MeV to 1 TeV (0.01-1000 GeV). Calculations were performed using Monte Carlo transport code MCNPX 2.7.C and BodyBuilder TM 1.3 anthropomorphic phantoms modified to allow calculation of effective dose using tissues and tissue weighting factors from either the 1990 or 2007 recommendations of the International Commission on Radiological Protection (ICRP), and gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. At 15 of the 19 energies for which coefficients for effective dose were calculated, coefficients based on ICRP 2007 and 1990 recommendations differed by less than 2%. The greatest difference, 62%, occurred at 100 MeV. Published by Oxford Univ. Press on behalf of the U.S. Government 2010. (authors)
International Nuclear Information System (INIS)
Copeland, K.; Parker, D. E.; Friedberg, W.
2011-01-01
Conversion coefficients were calculated for fluence-to-absorbed dose, fluence-to-equivalent dose, fluence-to-effective dose and fluence-to-gray equivalent for isotropic exposure of an adult female and an adult male to deuterons ( 2 H + ) in the energy range 10 MeV -1 TeV (0.01-1000 GeV). Coefficients were calculated using the Monte Carlo transport code MCNPX 2.7.C and BodyBuilder TM 1.3 anthropomorphic phantoms. Phantoms were modified to allow calculation of the effective dose to a Reference Person using tissues and tissue weighting factors from 1990 and 2007 recommendations of the International Commission on Radiological Protection (ICRP) and gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. Coefficients for the equivalent and effective dose incorporated a radiation weighting factor of 2. At 15 of 19 energies for which coefficients for the effective dose were calculated, coefficients based on ICRP 1990 and 2007 recommendations differed by < 3 %. The greatest difference, 47 %, occurred at 30 MeV. (authors)
International Nuclear Information System (INIS)
Murata, Isao; Mori, Takamasa; Nakagawa, Masayuki; Shirai, Hiroshi.
1996-03-01
High Temperature Gas-cooled Reactors (HTGRs) employ spherical fuels named coated fuel particles (CFPs) consisting of a microsphere of low enriched UO 2 with coating layers in order to prevent FP release. There exist many spherical fuels distributed randomly in the cores. Therefore, the nuclear design of HTGRs is generally performed on the basis of the multigroup approximation using a diffusion code, S N transport code or group-wise Monte Carlo code. This report summarizes a Monte Carlo hard sphere packing simulation code to simulate the packing of equal hard spheres and evaluate the necessary probability distribution of them, which is used for the application of the new Monte Carlo calculation method developed to treat randomly distributed spherical fuels with the continuous energy Monte Carlo method. By using this code, obtained are the various statistical values, namely Radial Distribution Function (RDF), Nearest Neighbor Distribution (NND), 2-dimensional RDF and so on, for random packing as well as ordered close packing of FCC and BCC. (author)
International Nuclear Information System (INIS)
Maconald, J.L.; Cashwell, E.D.
1978-09-01
The techniques of learning theory and pattern recognition are used to learn splitting surface locations for the Monte Carlo neutron transport code MCN. A study is performed to determine default values for several pattern recognition and learning parameters. The modified MCN code is used to reduce computer cost for several nontrivial example problems
International Nuclear Information System (INIS)
Ford, R.L.; Nelson, W.R.
1978-06-01
A code to simulate almost any electron--photon transport problem conceivable is described. The report begins with a lengthy historical introduction and a description of the shower generation process. Then the detailed physics of the shower processes and the methods used to simulate them are presented. Ideas of sampling theory, transport techniques, particle interactions in general, and programing details are discussed. Next, EGS calculations and various experiments and other Monte Carlo results are compared. The remainder of the report consists of user manuals for EGS, PEGS, and TESTSR codes; options, input specifications, and typical output are included. 38 figures, 12 tables
Development of fast and accurate Monte Carlo code MVP
International Nuclear Information System (INIS)
Mori, Takamasa
2001-01-01
The development work of fast and accurate Monte Carlo code MVP has started at JAERI in late 80s. From the beginning, the code was designed to utilize vector supercomputers and achieved higher computation speed by a factor of 10 or more compared with conventional codes. In 1994, the first version of MVP was released together with cross section libraries based on JENDL-3.1 and JENDL-3.2. In 1996, minor revision was made by adding several functions such as treatments of ENDF-B6 file 6 data, time dependent problem, and so on. Since 1996, several works have been carried out for the next version of MVP. The main works are (1) the development of continuous energy Monte Carlo burn-up calculation code MVP-BURN, (2) the development of a system to generate cross section libraries at arbitrary temperature, and (3) the study on error estimations and their biases in Monte Carlo eigenvalue calculations. This paper summarizes the main features of MVP, results of recent studies and future plans for MVP. (author)
CTCN: Colloid transport code -- nuclear
International Nuclear Information System (INIS)
Jain, R.
1993-01-01
This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential-algebraic equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential-algebraic systems
On the use of SERPENT Monte Carlo code to generate few group diffusion constants
Energy Technology Data Exchange (ETDEWEB)
Piovezan, Pamela, E-mail: pamela.piovezan@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Carluccio, Thiago; Domingos, Douglas Borges; Rossi, Pedro Russo; Mura, Luiz Felipe, E-mail: fermium@cietec.org.b, E-mail: thiagoc@ipen.b [Fermium Tecnologia Nuclear, Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2011-07-01
The accuracy of diffusion reactor codes strongly depends on the quality of the groups constants processing. For many years, the generation of such constants was based on 1-D infinity cell transport calculations. Some developments using collision probability or the method of characteristics allow, nowadays, 2-D assembly group constants calculations. However, these 1-D and 2-D codes how some limitations as , for example, on complex geometries and in the neighborhood of heavy absorbers. On the other hand, since Monte Carlos (MC) codes provide accurate neutro flux distributions, the possibility of using these solutions to provide group constants to full-core reactor diffusion simulators has been recently investigated, especially for the cases in which the geometry and reactor types are beyond the capability of the conventional deterministic lattice codes. The two greatest difficulties on the use of MC codes to group constant generation are the computational costs and the methodological incompatibility between analog MC particle transport simulation and deterministic transport methods based in several approximations. The SERPENT code is a 3-D continuous energy MC transport code with built-in burnup capability that was specially optimized to generate these group constants. In this work, we present the preliminary results of using the SERPENT MC code to generate 3-D two-group diffusion constants for a PWR like assembly. These constants were used in the CITATION diffusion code to investigate the effects of the MC group constants determination on the neutron multiplication factor diffusion estimate. (author)
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.
Computer codes in nuclear safety, radiation transport and dosimetry
International Nuclear Information System (INIS)
Bordy, J.M.; Kodeli, I.; Menard, St.; Bouchet, J.L.; Renard, F.; Martin, E.; Blazy, L.; Voros, S.; Bochud, F.; Laedermann, J.P.; Beaugelin, K.; Makovicka, L.; Quiot, A.; Vermeersch, F.; Roche, H.; Perrin, M.C.; Laye, F.; Bardies, M.; Struelens, L.; Vanhavere, F.; Gschwind, R.; Fernandez, F.; Quesne, B.; Fritsch, P.; Lamart, St.; Crovisier, Ph.; Leservot, A.; Antoni, R.; Huet, Ch.; Thiam, Ch.; Donadille, L.; Monfort, M.; Diop, Ch.; Ricard, M.
2006-01-01
The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations
Energy Technology Data Exchange (ETDEWEB)
Liang, Jingang; Wang, Kan; Qiu, Yishu [Dept. of Engineering Physics, LiuQing Building, Tsinghua University, Beijing (China); Chai, Xiao Ming; Qiang, Sheng Long [Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu (China)
2016-06-15
Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC) codes in accomplishing pin-wise three-dimensional (3D) full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC) code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.
Particle-transport simulation with the Monte Carlo method
International Nuclear Information System (INIS)
Carter, L.L.; Cashwell, E.D.
1975-01-01
Attention is focused on the application of the Monte Carlo method to particle transport problems, with emphasis on neutron and photon transport. Topics covered include sampling methods, mathematical prescriptions for simulating particle transport, mechanics of simulating particle transport, neutron transport, and photon transport. A literature survey of 204 references is included. (GMT)
Burnup calculation methodology in the serpent 2 Monte Carlo code
International Nuclear Information System (INIS)
Leppaenen, J.; Isotalo, A.
2012-01-01
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)
On-the-fly doppler broadening for Monte Carlo codes
International Nuclear Information System (INIS)
Yesilyurt, G.; Martin, W. R.; Brown, F. B.
2009-01-01
A methodology to allow on-the-fly Doppler broadening of neutron cross sections for use in Monte Carlo codes has been developed. The Monte Carlo code only needs to store 0 K cross sections for each isotope and the method will broaden the 0 K cross sections for any isotope in the library to any temperature in the range 77 K-3200 K. The methodology is based on a combination of Taylor series expansions and asymptotic series expansions. The type of series representation was determined by investigating the temperature dependence of U3o8 resonance cross sections in three regions: near the resonance peaks, mid-resonance, and the resonance wings. The coefficients for these series expansions were determined by a regression over the energy and temperature range of interest. Since the resonance parameters are a function of the neutron energy and target nuclide, the ψ and χ functions in the Adler-Adler multi-level resonance model can be represented by series expansions in temperature only, allowing the least number of terms to approximate the temperature dependent cross sections within a given accuracy. The comparison of the broadened cross sections using this methodology with the NJOY cross sections was excellent over the entire temperature range (77 K-3200 K) and energy range. A Monte Carlo code was implemented to apply the combined regression model and used to estimate the additional computing cost which was found to be less than <1%. (authors)
Portable LQCD Monte Carlo code using OpenACC
Bonati, Claudio; Calore, Enrico; Coscetti, Simone; D'Elia, Massimo; Mesiti, Michele; Negro, Francesco; Fabio Schifano, Sebastiano; Silvi, Giorgio; Tripiccione, Raffaele
2018-03-01
Varying from multi-core CPU processors to many-core GPUs, the present scenario of HPC architectures is extremely heterogeneous. In this context, code portability is increasingly important for easy maintainability of applications; this is relevant in scientific computing where code changes are numerous and frequent. In this talk we present the design and optimization of a state-of-the-art production level LQCD Monte Carlo application, using the OpenACC directives model. OpenACC aims to abstract parallel programming to a descriptive level, where programmers do not need to specify the mapping of the code on the target machine. We describe the OpenACC implementation and show that the same code is able to target different architectures, including state-of-the-art CPUs and GPUs.
Development and validation of ALEPH Monte Carlo burn-up code
International Nuclear Information System (INIS)
Stankovskiy, A.; Van den Eynde, G.; Vidmar, T.
2011-01-01
The Monte-Carlo burn-up code ALEPH is being developed in SCK-CEN since 2004. Belonging to the category of shells coupling Monte Carlo transport (MCNP or MCNPX) and 'deterministic' depletion codes (ORIGEN-2.2), ALEPH possess some unique features that distinguish it from other codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. Recent improvements of ALEPH concern full implementation of general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII, JENDL-3.3). The upgraded version of the code is capable to treat isomeric branching ratios, neutron induced fission product yields, spontaneous fission yields and energy release per fission recorded in ENDF-formatted data files. The alternative algorithm for time evolution of nuclide concentrations is added. A predictor-corrector mechanism and the calculation of nuclear heating are available as well. The validation of the code on REBUS experimental programme results has been performed. The upgraded version of ALEPH has shown better agreement with measured data than other codes, including previous version of ALEPH. (authors)
Effects of physics change in Monte Carlo code on electron pencil beam dose distributions
International Nuclear Information System (INIS)
Toutaoui, Abdelkader; Khelassi-Toutaoui, Nadia; Brahimi, Zakia; Chami, Ahmed Chafik
2012-01-01
Pencil beam algorithms used in computerized electron beam dose planning are usually described using the small angle multiple scattering theory. Alternatively, the pencil beams can be generated by Monte Carlo simulation of electron transport. In a previous work, the 4th version of the Electron Gamma Shower (EGS) Monte Carlo code was used to obtain dose distributions from monoenergetic electron pencil beam, with incident energy between 1 MeV and 50 MeV, interacting at the surface of a large cylindrical homogeneous water phantom. In 2000, a new version of this Monte Carlo code has been made available by the National Research Council of Canada (NRC), which includes various improvements in its electron-transport algorithms. In the present work, we were interested to see if the new physics in this version produces pencil beam dose distributions very different from those calculated with oldest one. The purpose of this study is to quantify as well as to understand these differences. We have compared a series of pencil beam dose distributions scored in cylindrical geometry, for electron energies between 1 MeV and 50 MeV calculated with two versions of the Electron Gamma Shower Monte Carlo Code. Data calculated and compared include isodose distributions, radial dose distributions and fractions of energy deposition. Our results for radial dose distributions show agreement within 10% between doses calculated by the two codes for voxels closer to the pencil beam central axis, while the differences are up to 30% for longer distances. For fractions of energy deposition, the results of the EGS4 are in good agreement (within 2%) with those calculated by EGSnrc at shallow depths for all energies, whereas a slightly worse agreement (15%) is observed at deeper distances. These differences may be mainly attributed to the different multiple scattering for electron transport adopted in these two codes and the inclusion of spin effect, which produces an increase of the effective range of
Organization of cross-section data in the Monte Carlo code SPARTAN
International Nuclear Information System (INIS)
Bending, R.C.
1974-01-01
The Monte Carlo code SPARTAN is a general-purpose code intended for neutron or gamma transport calculations. The code is designed to accept physics data from a number of external libraries (which may be used singly or in combination) and to use this data with as little alteration as possible. Data obtained from one or several libraries is placed in an interface file on magnetic tape or disk, using a general hierarchical structure which allows particular data items to be assessed in a straightforward way. The interface file, with or without additional data from cards, is regarded as a data source for the main Monte Carlo calculation. A summary of the functional forms, sampling distributions, and particle interaction laws which are available at present, and some of the mathematical methods used are described. 5 references. (U.S.)
Recent developments of JAEA's Monte Carlo Code MVP for reactor physics applications
International Nuclear Information System (INIS)
Nagaya, Y.; Okumura, K.; Mori, T.
2013-01-01
MVP is a general-purpose continuous-energy Monte Carlo code for neutron and photon transport calculations that has been developed since the late 1980's at Japan Atomic Energy Agency (JAEA, formerly JAERI). The MVP code is designed for nuclear reactor applications such as reactor core design/analysis, criticality safety and reactor shielding. This paper describes the MVP code and present its latest developments. Among the new capabilities of MVP we find: -) the perturbation method has been implemented for the change in k(eff); -) the eigenvalue calculations can be performed with an explicit treatment of delayed neutrons in which their fission spectra are taken into account; -) the capability of tallying the scattering matrix (group-to-group scattering cross sections); -) the implementation of an exact model for resonance elastic scattering; and -) a Monte Carlo perturbation technique is used to calculate reactor kinetics parameters
ERSN-OpenMC, a Java-based GUI for OpenMC Monte Carlo code
Directory of Open Access Journals (Sweden)
Jaafar EL Bakkali
2016-07-01
Full Text Available OpenMC is a new Monte Carlo transport particle simulation code focused on solving two types of neutronic problems mainly the k-eigenvalue criticality fission source problems and external fixed fission source problems. OpenMC does not have any Graphical User Interface and the creation of one is provided by our java-based application named ERSN-OpenMC. The main feature of this application is to provide to the users an easy-to-use and flexible graphical interface to build better and faster simulations, with less effort and great reliability. Additionally, this graphical tool was developed with several features, as the ability to automate the building process of OpenMC code and related libraries as well as the users are given the freedom to customize their installation of this Monte Carlo code. A full description of the ERSN-OpenMC application is presented in this paper.
TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticality studies
International Nuclear Information System (INIS)
Ermumcu, G.; Gonnord, J.; Nimal, J.C.
1980-01-01
TRIMARAN is developed for safety analysis of nuclear components containing fissionable materials: shipping casks, storage and cooling pools, manufacture and reprocessing plants. It solves the transport equation by Monte Carlo method, in general three dimensional geometry with multigroup P1 approximation. A special representation of cross sections and numbers has been developed in order to reduce considerably the computing cost and allow this three dimensional code to compete with standard numerical program used in parametric studies
TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticallity studies
International Nuclear Information System (INIS)
Ermuncu, G.; Gonnord, J.; Nimal, J.C.
1980-04-01
TRIMARAN is developed for safety analysis of nuclar components containing fissionnable materials: shipping casks, storage and cooling pools, manufacture and reprocessing plants. It solves the transport equation by Monte Carlo method in general three dimensional geometry with multigroup P1 approximation. A special representation of cross sections and numbers has been developed in order to reduce considerably the computing cost and allow this three dimensional code to compete with standard numerical program used in parametric studies
Comparison of EGS4 and MCNP Monte Carlo codes when calculating radiotherapy depth doses.
Love, P A; Lewis, D G; Al-Affan, I A; Smith, C W
1998-05-01
The Monte Carlo codes EGS4 and MCNP have been compared when calculating radiotherapy depth doses in water. The aims of the work were to study (i) the differences between calculated depth doses in water for a range of monoenergetic photon energies and (ii) the relative efficiency of the two codes for different electron transport energy cut-offs. The depth doses from the two codes agree with each other within the statistical uncertainties of the calculations (1-2%). The relative depth doses also agree with data tabulated in the British Journal of Radiology Supplement 25. A discrepancy in the dose build-up region may by attributed to the different electron transport algorithims used by EGS4 and MCNP. This discrepancy is considerably reduced when the improved electron transport routines are used in the latest (4B) version of MCNP. Timing calculations show that EGS4 is at least 50% faster than MCNP for the geometries used in the simulations.
Validation and verification of the ORNL Monte Carlo codes for nuclear safety analysis
International Nuclear Information System (INIS)
Emmett, M.B.
1993-01-01
The process of ensuring the quality of computer codes can be very time consuming and expensive. The Oak Ridge National Laboratory (ORNL) Monte Carlo codes all predate the existence of quality assurance (QA) standards and configuration control. The number of person-years and the amount of money spent on code development make it impossible to adhere strictly to all the current requirements. At ORNL, the Nuclear Engineering Applications Section of the Computing Applications Division is responsible for the development, maintenance, and application of the Monte Carlo codes MORSE and KENO. The KENO code is used for doing criticality analyses; the MORSE code, which has two official versions, CGA and SGC, is used for radiation transport analyses. Because KENO and MORSE were very thoroughly checked out over the many years of extensive use both in the United States and in the international community, the existing codes were open-quotes baselined.close quotes This means that the versions existing at the time the original configuration plan is written are considered to be validated and verified code systems based on the established experience with them
Improving system modeling accuracy with Monte Carlo codes
International Nuclear Information System (INIS)
Johnson, A.S.
1996-01-01
The use of computer codes based on Monte Carlo methods to perform criticality calculations has become common-place. Although results frequently published in the literature report calculated k eff values to four decimal places, people who use the codes in their everyday work say that they only believe the first two decimal places of any result. The lack of confidence in the computed k eff values may be due to the tendency of the reported standard deviation to underestimate errors associated with the Monte Carlo process. The standard deviation as reported by the codes is the standard deviation of the mean of the k eff values for individual generations in the computer simulation, not the standard deviation of the computed k eff value compared with the physical system. A more subtle problem with the standard deviation of the mean as reported by the codes is that all the k eff values from the separate generations are not statistically independent since the k eff of a given generation is a function of k eff of the previous generation, which is ultimately based on the starting source. To produce a standard deviation that is more representative of the physical system, statistically independent values of k eff are needed
OPAL reactor calculations using the Monte Carlo code serpent
Energy Technology Data Exchange (ETDEWEB)
Ferraro, Diego; Villarino, Eduardo [Nuclear Engineering Dept., INVAP S.E., Rio Negro (Argentina)
2012-03-15
In the present work the Monte Carlo cell code developed by VTT Serpent v1.1.14 is used to model the MTR fuel assemblies (FA) and control rods (CR) from OPAL (Open Pool Australian Light-water) reactor in order to obtain few-group constants with burnup dependence to be used in the already developed reactor core models. These core calculations are performed using CITVAP 3-D diffusion code, which is well-known reactor code based on CITATION. Subsequently the results are compared with those obtained by the deterministic calculation line used by INVAP, which uses the Collision Probability Condor cell-code to obtain few-group constants. Finally the results are compared with the experimental data obtained from the reactor information for several operation cycles. As a result several evaluations are performed, including a code to code cell comparison at cell and core level and calculation-experiment comparison at core level in order to evaluate the Serpent code actual capabilities. (author)
Exponential convergence on a continuous Monte Carlo transport problem
International Nuclear Information System (INIS)
Booth, T.E.
1997-01-01
For more than a decade, it has been known that exponential convergence on discrete transport problems was possible using adaptive Monte Carlo techniques. An adaptive Monte Carlo method that empirically produces exponential convergence on a simple continuous transport problem is described
Development of Monte Carlo decay gamma-ray transport calculation system
Energy Technology Data Exchange (ETDEWEB)
Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Kawasaki, Nobuo [Fujitsu Ltd., Tokyo (Japan); Kume, Etsuo [Japan Atomic Energy Research Inst., Center for Promotion of Computational Science and Engineering, Tokai, Ibaraki (Japan)
2001-06-01
In the DT fusion reactor, it is critical concern to evaluate the decay gamma-ray biological dose rates after the reactor shutdown exactly. In order to evaluate the decay gamma-ray biological dose rates exactly, three dimensional Monte Carlo decay gamma-ray transport calculation system have been developed by connecting the three dimensional Monte Carlo particle transport calculation code and the induced activity calculation code. The developed calculation system consists of the following four functions. (1) The operational neutron flux distribution is calculated by the three dimensional Monte Carlo particle transport calculation code. (2) The induced activities are calculated by the induced activity calculation code. (3) The decay gamma-ray source distribution is obtained from the induced activities. (4) The decay gamma-rays are generated by using the decay gamma-ray source distribution, and the decay gamma-ray transport calculation is conducted by the three dimensional Monte Carlo particle transport calculation code. In order to reduce the calculation time drastically, a biasing system for the decay gamma-ray source distribution has been developed, and the function is also included in the present system. In this paper, the outline and the detail of the system, and the execution example are reported. The evaluation for the effect of the biasing system is also reported. (author)
2014-03-27
Vehicle Code System (VCS), the Monte Carlo Adjoint SHielding (MASH), and the Monte Carlo n- Particle ( MCNP ) code. Of the three, the oldest and still most...widely utilized radiation transport code is MCNP . First created at Los Alamos National Laboratory (LANL) in 1957, the code simulated neutral...particle types, and previous versions of MCNP were repeatedly validated using both simple and complex 10 geometries [12, 13]. Much greater discussion and
Comparative Criticality Analysis of Two Monte Carlo Codes on Centrifugal Atomizer: MCNPS and SCALE
International Nuclear Information System (INIS)
Kang, H-S; Jang, M-S; Kim, S-R; Park, J-M; Kim, K-N
2015-01-01
There are two well-known Monte Carlo codes for criticality analysis, MCNP5 and SCALE. MCNP5 is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical system as a main analysis code. SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. SCALE was conceived and funded by US NRC to perform standardized computer analysis for licensing evaluation and is used widely in the world. We performed a validation test of MCNP5 and a comparative analysis of Monte Carlo codes, MCNP5 and SCALE, in terms of the critical analysis of centrifugal atomizer. In the criticality analysis using MCNP5 code, we obtained the statistically reliable results by using a large number of source histories per cycle and performing of uncertainty analysis
Comparative Criticality Analysis of Two Monte Carlo Codes on Centrifugal Atomizer: MCNPS and SCALE
Energy Technology Data Exchange (ETDEWEB)
Kang, H-S; Jang, M-S; Kim, S-R [NESS, Daejeon (Korea, Republic of); Park, J-M; Kim, K-N [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2015-10-15
There are two well-known Monte Carlo codes for criticality analysis, MCNP5 and SCALE. MCNP5 is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical system as a main analysis code. SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. SCALE was conceived and funded by US NRC to perform standardized computer analysis for licensing evaluation and is used widely in the world. We performed a validation test of MCNP5 and a comparative analysis of Monte Carlo codes, MCNP5 and SCALE, in terms of the critical analysis of centrifugal atomizer. In the criticality analysis using MCNP5 code, we obtained the statistically reliable results by using a large number of source histories per cycle and performing of uncertainty analysis.
International Nuclear Information System (INIS)
Yamaguchi, Yasuhiro
1991-01-01
The present report describes a computer code DEEP which calculates the organ dose equivalents and the effective dose equivalent for external photon exposure by the Monte Carlo method. MORSE-CG, Monte Carlo radiation transport code, is incorporated into the DEEP code to simulate photon transport phenomena in and around a human body. The code treats an anthropomorphic phantom represented by mathematical formulae and user has a choice for the phantom sex: male, female and unisex. The phantom can wear personal dosimeters on it and user can specify their location and dimension. This document includes instruction and sample problem for the code as well as the general description of dose calculation, human phantom and computer code. (author)
Statistics of Monte Carlo methods used in radiation transport calculation
International Nuclear Information System (INIS)
Datta, D.
2009-01-01
Radiation transport calculation can be carried out by using either deterministic or statistical methods. Radiation transport calculation based on statistical methods is basic theme of the Monte Carlo methods. The aim of this lecture is to describe the fundamental statistics required to build the foundations of Monte Carlo technique for radiation transport calculation. Lecture note is organized in the following way. Section (1) will describe the introduction of Basic Monte Carlo and its classification towards the respective field. Section (2) will describe the random sampling methods, a key component of Monte Carlo radiation transport calculation, Section (3) will provide the statistical uncertainty of Monte Carlo estimates, Section (4) will describe in brief the importance of variance reduction techniques while sampling particles such as photon, or neutron in the process of radiation transport
Extensions of the MCNP5 and TRIPOLI4 Monte Carlo codes for transient reactor analysis
International Nuclear Information System (INIS)
Hoogenboom, J.E.
2013-01-01
To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branch-less collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires the coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3*3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3*3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail. (authors)
Introduction to the simulation with MCNP Monte Carlo code and its applications in Medical Physics
International Nuclear Information System (INIS)
Parreno Z, F.; Paucar J, R.; Picon C, C.
1998-01-01
The simulation by Monte Carlo is tool which Medical Physics counts with it for the development of its research, the interest by this tool is growing, as we may observe in the main scientific journals for the years 1995-1997 where more than 27 % of the papers treat over Monte Carlo and/or its applications in the radiation transport.In the Peruvian Institute of Nuclear Energy we are implementing and making use of the MCNP4 and EGS4 codes. In this work are presented the general features of the Monte Carlo method and its more useful applications in Medical Physics. Likewise, it is made a simulation of the calculation of isodose curves in an interstitial treatment with Ir-192 wires in a mammary gland carcinoma. (Author)
Criticality qualification of a new Monte Carlo code for reactor core analysis
International Nuclear Information System (INIS)
Catsaros, N.; Gaveau, B.; Jaekel, M.; Maillard, J.; Maurel, G.; Savva, P.; Silva, J.; Varvayanni, M.; Zisis, Th.
2009-01-01
In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal-hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the 'Accelerator part' of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the 'Reactor part' of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.
Efficient data management techniques implemented in the Karlsruhe Monte Carlo code KAMCCO
International Nuclear Information System (INIS)
Arnecke, G.; Borgwaldt, H.; Brandl, V.; Lalovic, M.
1974-01-01
The Karlsruhe Monte Carlo Code KAMCCO is a forward neutron transport code with an eigenfunction and a fixed source option, including time-dependence. A continuous energy model is combined with a detailed representation of neutron cross sections, based on linear interpolation, Breit-Wigner resonances and probability tables. All input is processed into densely packed, dynamically addressed parameter fields and networks of pointers (addresses). Estimation routines are decoupled from random walk and analyze a storage region with sample records. This technique leads to fast execution with moderate storage requirements and without any I/O-operations except in the input and output stages. 7 references. (U.S.)
Progress in nuclear well logging modeling using deterministic transport codes
International Nuclear Information System (INIS)
Kodeli, I.; Aldama, D.L.; Maucec, M.; Trkov, A.
2002-01-01
Further studies in continuation of the work presented in 2001 in Portoroz were performed in order to study and improve the performances, precission and domain of application of the deterministic transport codes with respect to the oil well logging analysis. These codes are in particular expected to complement the Monte Carlo solutions, since they can provide a detailed particle flux distribution in the whole geometry in a very reasonable CPU time. Real-time calculation can be envisaged. The performances of deterministic transport methods were compared to those of the Monte Carlo method. IRTMBA generic benchmark was analysed using the codes MCNP-4C and DORT/TORT. Centric as well as excentric casings were considered using 14 MeV point neutron source and NaI scintillation detectors. Neutron and gamma spectra were compared at two detector positions.(author)
Criticality benchmarks for COG: A new point-wise Monte Carlo code
International Nuclear Information System (INIS)
Alesso, H.P.; Pearson, J.; Choi, J.S.
1989-01-01
COG is a new point-wise Monte Carlo code being developed and tested at LLNL for the Cray computer. It solves the Boltzmann equation for the transport of neutrons, photons, and (in future versions) charged particles. Techniques included in the code for modifying the random walk of particles make COG most suitable for solving deep-penetration (shielding) problems. However, its point-wise cross-sections also make it effective for a wide variety of criticality problems. COG has some similarities to a number of other computer codes used in the shielding and criticality community. These include the Lawrence Livermore National Laboratory (LLNL) codes TART and ALICE, the Los Alamos National Laboratory code MCNP, the Oak Ridge National Laboratory codes 05R, 06R, KENO, and MORSE, the SACLAY code TRIPOLI, and the MAGI code SAM. Each code is a little different in its geometry input and its random-walk modification options. Validating COG consists in part of running benchmark calculations against critical experiments as well as other codes. The objective of this paper is to present calculational results of a variety of critical benchmark experiments using COG, and to present the resulting code bias. Numerous benchmark calculations have been completed for a wide variety of critical experiments which generally involve both simple and complex physical problems. The COG results, which they report in this paper, have been excellent
Modeling Dynamic Objects in Monte Carlo Particle Transport Calculations
International Nuclear Information System (INIS)
Yegin, G.
2008-01-01
In this study, the Multi-Geometry geometry modeling technique was improved in order to handle moving objects in a Monte Carlo particle transport calculation. In the Multi-Geometry technique, the geometry is a superposition of objects not surfaces. By using this feature, we developed a new algorithm which allows a user to make enable or disable geometry elements during particle transport. A disabled object can be ignored at a certain stage of a calculation and switching among identical copies of the same object located adjacent poins during a particle simulation corresponds to the movement of that object in space. We called this powerfull feature as Dynamic Multi-Geometry technique (DMG) which is used for the first time in Brachy Dose Monte Carlo code to simulate HDR brachytherapy treatment systems. Our results showed that having disabled objects in a geometry does not effect calculated dose values. This technique is also suitable to be used in other areas such as IMRT treatment planning systems
Computed radiography simulation using the Monte Carlo code MCNPX
International Nuclear Information System (INIS)
Correa, S.C.A.; Souza, E.M.; Silva, A.X.; Lopes, R.T.
2009-01-01
Simulating x-ray images has been of great interest in recent years as it makes possible an analysis of how x-ray images are affected owing to relevant operating parameters. In this paper, a procedure for simulating computed radiographic images using the Monte Carlo code MCNPX is proposed. The sensitivity curve of the BaFBr image plate detector as well as the characteristic noise of a 16-bit computed radiography system were considered during the methodology's development. The results obtained confirm that the proposed procedure for simulating computed radiographic images is satisfactory, as it allows obtaining results comparable with experimental data. (author)
Computed radiography simulation using the Monte Carlo code MCNPX
Energy Technology Data Exchange (ETDEWEB)
Correa, S.C.A. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Centro Universitario Estadual da Zona Oeste (CCMAT)/UEZO, Av. Manuel Caldeira de Alvarenga, 1203, Campo Grande, 23070-200, Rio de Janeiro, RJ (Brazil); Souza, E.M. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Silva, A.X., E-mail: ademir@con.ufrj.b [PEN/COPPE-DNC/Poli CT, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Cassiano, D.H. [Instituto de Radioprotecao e Dosimetria/CNEN Av. Salvador Allende, s/n, Recreio, 22780-160, Rio de Janeiro, RJ (Brazil); Lopes, R.T. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil)
2010-09-15
Simulating X-ray images has been of great interest in recent years as it makes possible an analysis of how X-ray images are affected owing to relevant operating parameters. In this paper, a procedure for simulating computed radiographic images using the Monte Carlo code MCNPX is proposed. The sensitivity curve of the BaFBr image plate detector as well as the characteristic noise of a 16-bit computed radiography system were considered during the methodology's development. The results obtained confirm that the proposed procedure for simulating computed radiographic images is satisfactory, as it allows obtaining results comparable with experimental data.
Solution weighting for the SAND-II Monte Carlo code
International Nuclear Information System (INIS)
Oster, C.A.; McElroy, W.N.; Simons, R.L.; Lippincott, E.P.; Odette, G.R.
1976-01-01
Modifications to the SAND-II Error Analysis Monte Carlo code to include solution weighting based on input data uncertainties have been made and are discussed together with background information on the SAND-II algorithm. The new procedure permits input data having smaller uncertainties to have a greater influence on the solution spectrum than do the data having larger uncertainties. The results of an indepth study to find a practical procedure and the first results of its application to three important Interlaboratory LMFBR Reaction Rate (ILRR) program benchmark spectra (CFRMF, ΣΣ, and 235 U fission) are discussed
International Nuclear Information System (INIS)
Nishida, Takahiko; Horikami, Kunihiko; Suzuki, Tadakazu; Nakahara, Yasuaki; Taji, Yukichi
1975-09-01
The coarse-mesh rebalancing technique is introduced into the general-purpose neutron and gamma-ray Monte Carlo transport code MORSE, to accelerate the convergence rate of the iteration process for eigenvalue calculation in a nuclear reactor system. Two subroutines are thus attached to the code. One is bookkeeping routine 'COARSE' for obtaining the quantities related with the neutron balance in each coarse mesh cell, such as the number of neutrons absorbed in the cell, from random walks of neutrons in a batch. The other is rebalance factor calculation routine 'REBAL' for obtaining the scaling factor whereby the neutron flux in the cell is multiplied to attain the neutron balance. The two subroutines and algorithm of the coarse mesh rebalancing acceleration in a Monte Carlo game are described. (auth.)
Review of the Monte Carlo and deterministic codes in radiation protection and dosimetry
International Nuclear Information System (INIS)
Tagziria, H.
2000-02-01
Modelling a physical system can be carried out either stochastically or deterministically. An example of the former method is the Monte Carlo technique, in which statistically approximate methods are applied to exact models. No transport equation is solved as individual particles are simulated and some specific aspect (tally) of their average behaviour is recorded. The average behaviour of the physical system is then inferred using the central limit theorem. In contrast, deterministic codes use mathematically exact methods that are applied to approximate models to solve the transport equation for the average particle behaviour. The physical system is subdivided in boxes in the phase-space system and particles are followed from one box to the next. The smaller the boxes the better the approximations become. Although the Monte Carlo method has been used for centuries, its more recent manifestation has really emerged from the Manhattan project of the Word War II. Its invention is thought to be mainly due to Metropolis, Ulah (through his interest in poker), Fermi, von Neuman and Richtmeyer. Over the last 20 years or so, the Monte Carlo technique has become a powerful tool in radiation transport. This is due to users taking full advantage of richer cross section data, more powerful computers and Monte Carlo techniques for radiation transport, with high quality physics and better known source spectra. This method is a common sense approach to radiation transport and its success and popularity is quite often also due to necessity, because measurements are not always possible or affordable. In the Monte Carlo method, which is inherently realistic because nature is statistical, a more detailed physics is made possible by isolation of events while rather elaborate geometries can be modelled. Provided that the physics is correct, a simulation is exactly analogous to an experimenter counting particles. In contrast to the deterministic approach, however, a disadvantage of the
International Nuclear Information System (INIS)
Bourauel, Peter; Nabbi, Rahim; Biel, Wolfgang; Forrest, Robin
2009-01-01
The MCNP 3D Monte Carlo computer code is used not only for criticality calculations of nuclear systems but also to simulate transports of radiation and particles. The findings so obtained about neutron flux distribution and the associated spectra allow information about materials activation, nuclear heating, and radiation damage to be obtained by means of activation codes such as FISPACT. The stochastic character of particle and radiation transport processes normally links findings to the materials cells making up the geometry model of MCNP. Where high spatial resolution is required for the activation calculations with FISPACT, fine segmentation of the MCNP geometry becomes compulsory, which implies considerable expense for the modeling process. For this reason, an alternative simulation technique has been developed in an effort to automate and optimize data transfer between MCNP and FISPACT. (orig.)
Development of a space radiation Monte Carlo computer simulation based on the FLUKA and ROOT codes
Pinsky, L; Ferrari, A; Sala, P; Carminati, F; Brun, R
2001-01-01
This NASA funded project is proceeding to develop a Monte Carlo-based computer simulation of the radiation environment in space. With actual funding only initially in place at the end of May 2000, the study is still in the early stage of development. The general tasks have been identified and personnel have been selected. The code to be assembled will be based upon two major existing software packages. The radiation transport simulation will be accomplished by updating the FLUKA Monte Carlo program, and the user interface will employ the ROOT software being developed at CERN. The end-product will be a Monte Carlo-based code which will complement the existing analytic codes such as BRYNTRN/HZETRN presently used by NASA to evaluate the effects of radiation shielding in space. The planned code will possess the ability to evaluate the radiation environment for spacecraft and habitats in Earth orbit, in interplanetary space, on the lunar surface, or on a planetary surface such as Mars. Furthermore, it will be usef...
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.
International Nuclear Information System (INIS)
Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio.
1996-05-01
A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)
Energy Technology Data Exchange (ETDEWEB)
Mori, Takamasa; Nakagawa, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio
1996-05-01
A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author).
The three-dimensional Monte-Carlo code TRIPOLI-02
International Nuclear Information System (INIS)
Baur, A.; Bourdet, L.; Dejonghe, G.; Gonnord, J.; Monnier, A.; Nimal, J.C.; Vergnaud, T.
1980-04-01
TRIPOLI-2 solves the transport equation for neutrons or gamma rays in tridimensional geometrical configurations. TRIPOLI uses the Monte Carlo method. This method allows to treat exactly the geometrical configurations, the energy losses and the scattering laws. TRIPOLI 2 allows to treat the following problems: gamma transport problems, neutrons transport problems with fixed source (the problems can be time dependent or not), critical problems without fixed source and research of multiplication factor due to fissions, subcritical problems with fixed source and with multiplication by fission. These problems can be separate in two types. First type: shielding problems essentially with deep penetration and streaming through voids. Biasing technics are used to reduce the computing time. Second type: core problems for cell calculations or for small core calculations. In this case, it is necessary to have a fine representation of the cross sections. The thermalization is also treated exactly [fr
PHITS-a particle and heavy ion transport code system
International Nuclear Information System (INIS)
Niita, Koji; Sato, Tatsuhiko; Iwase, Hiroshi; Nose, Hiroyuki; Nakashima, Hiroshi; Sihver, Lembit
2006-01-01
The paper presents a summary of the recent development of the multi-purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS. In particular, we discuss in detail the development of two new models, JAM and JQMD, for high energy particle interactions, incorporated in PHITS, and show comparisons between model calculations and experiments for the validations of these models. The paper presents three applications of the code including spallation neutron source, heavy ion therapy and space radiation. The results and examples shown indicate PHITS has great ability of carrying out the radiation transport analysis of almost all particles including heavy ions within a wide energy range
Uncertainties associated with the use of the KENO Monte Carlo criticality codes
International Nuclear Information System (INIS)
Landers, N.F.; Petrie, L.M.
1989-01-01
The KENO multi-group Monte Carlo criticality codes have earned the reputation of being efficient, user friendly tools especially suited for the analysis of situations commonly encountered in the storage and transportation of fissile materials. Throughout their twenty years of service, a continuing effort has been made to maintain and improve these codes to meet the needs of the nuclear criticality safety community. Foremost among these needs is the knowledge of how to utilize the results safely and effectively. Therefore it is important that code users be aware of uncertainties that may affect their results. These uncertainties originate from approximations in the problem data, methods used to process cross sections, and assumptions, limitations and approximations within the criticality computer code itself. 6 refs., 8 figs., 1 tab
Parallel computing by Monte Carlo codes MVP/GMVP
International Nuclear Information System (INIS)
Nagaya, Yasunobu; Nakagawa, Masayuki; Mori, Takamasa
2001-01-01
General-purpose Monte Carlo codes MVP/GMVP are well-vectorized and thus enable us to perform high-speed Monte Carlo calculations. In order to achieve more speedups, we parallelized the codes on the different types of parallel computing platforms or by using a standard parallelization library MPI. The platforms used for benchmark calculations are a distributed-memory vector-parallel computer Fujitsu VPP500, a distributed-memory massively parallel computer Intel paragon and a distributed-memory scalar-parallel computer Hitachi SR2201, IBM SP2. As mentioned generally, linear speedup could be obtained for large-scale problems but parallelization efficiency decreased as the batch size per a processing element(PE) was smaller. It was also found that the statistical uncertainty for assembly powers was less than 0.1% by the PWR full-core calculation with more than 10 million histories and it took about 1.5 hours by massively parallel computing. (author)
Application of Monte Carlo codes to neutron dosimetry
International Nuclear Information System (INIS)
Prevo, C.T.
1982-01-01
In neutron dosimetry, calculations enable one to predict the response of a proposed dosimeter before effort is expended to design and fabricate the neutron instrument or dosimeter. The nature of these calculations requires the use of computer programs that implement mathematical models representing the transport of radiation through attenuating media. Numerical, and in some cases analytical, solutions of these models can be obtained by one of several calculational techniques. All of these techniques are either approximate solutions to the well-known Boltzmann equation or are based on kernels obtained from solutions to the equation. The Boltzmann equation is a precise mathematical description of neutron behavior in terms of position, energy, direction, and time. The solution of the transport equation represents the average value of the particle flux density. Integral forms of the transport equation are generally regarded as the formal basis for the Monte Carlo method, the results of which can in principle be made to approach the exact solution. This paper focuses on the Monte Carlo technique
Microwave transport in EBT distribution manifolds using Monte Carlo ray-tracing techniques
International Nuclear Information System (INIS)
Lillie, R.A.; White, T.L.; Gabriel, T.A.; Alsmiller, R.G. Jr.
1983-01-01
Ray tracing Monte Carlo calculations have been carried out using an existing Monte Carlo radiation transport code to obtain estimates of the microsave power exiting the torus coupling links in EPT microwave manifolds. The microwave power loss and polarization at surface reflections were accounted for by treating the microwaves as plane waves reflecting off plane surfaces. Agreement on the order of 10% was obtained between the measured and calculated output power distribution for an existing EBT-S toroidal manifold. A cost effective iterative procedure utilizing the Monte Carlo history data was implemented to predict design changes which could produce increased manifold efficiency and improved output power uniformity
Hybrid transport and diffusion modeling using electron thermal transport Monte Carlo SNB in DRACO
Chenhall, Jeffrey; Moses, Gregory
2017-10-01
The iSNB (implicit Schurtz Nicolai Busquet) multigroup diffusion electron thermal transport method is adapted into an Electron Thermal Transport Monte Carlo (ETTMC) transport method to better model angular and long mean free path non-local effects. Previously, the ETTMC model had been implemented in the 2D DRACO multiphysics code and found to produce consistent results with the iSNB method. Current work is focused on a hybridization of the computationally slower but higher fidelity ETTMC transport method with the computationally faster iSNB diffusion method in order to maximize computational efficiency. Furthermore, effects on the energy distribution of the heat flux divergence are studied. Work to date on the hybrid method will be presented. This work was supported by Sandia National Laboratories and the Univ. of Rochester Laboratory for Laser Energetics.
Monte Carlo impurity transport modeling in the DIII-D transport
International Nuclear Information System (INIS)
Evans, T.E.; Finkenthal, D.F.
1998-04-01
A description of the carbon transport and sputtering physics contained in the Monte Carlo Impurity (MCI) transport code is given. Examples of statistically significant carbon transport pathways are examined using MCI's unique tracking visualizer and a mechanism for enhanced carbon accumulation on the high field side of the divertor chamber is discussed. Comparisons between carbon emissions calculated with MCI and those measured in the DIII-D tokamak are described. Good qualitative agreement is found between 2D carbon emission patterns calculated with MCI and experimentally measured carbon patterns. While uncertainties in the sputtering physics, atomic data, and transport models have made quantitative comparisons with experiments more difficult, recent results using a physics based model for physical and chemical sputtering has yielded simulations with about 50% of the total carbon radiation measured in the divertor. These results and plans for future improvement in the physics models and atomic data are discussed
International Nuclear Information System (INIS)
Anwarul Islam, M.; Akramuzzaman, M.M.; Zakaria, G.A.
2012-01-01
Manufacturing of miniaturized high activity 192 Ir sources have been made a market preference in modern brachytherapy. The smaller dimensions of the sources are flexible for smaller diameter of the applicators and it is also suitable for interstitial implants. Presently, miniaturized 60 Co HDR sources have been made available with identical dimensions to those of 192 Ir sources. 60 Co sources have an advantage of longer half life while comparing with 192 Ir source. High dose rate brachytherapy sources with longer half life are logically pragmatic solution for developing country in economic point of view. This study is aimed to compare the TG-43U1 dosimetric parameters for new BEBIG 60 Co HDR and new microSelectron 192 Ir HDR sources. Dosimetric parameters are calculated using EGSnrc-based Monte Carlo simulation code accordance with the AAPM TG-43 formalism for microSelectron HDR 192 Ir v2 and new BEBIG 60 Co HDR sources. Air-kerma strength per unit source activity, calculated in dry air are 9.698x10 -8 ± 0.55% U Bq -1 and 3.039x10 -7 ± 0.41% U Bq -1 for the above mentioned two sources, respectively. The calculated dose rate constants per unit air-kerma strength in water medium are 1.116±0.12% cGy h -1 U -1 and 1.097±0.12% cGy h -1 U -1 , respectively, for the two sources. The values of radial dose function for distances up to 1 cm and more than 22 cm for BEBIG 60 Co HDR source are higher than that of other source. The anisotropic values are sharply increased to the longitudinal sides of the BEBIG 60 Co source and the rise is comparatively sharper than that of the other source. Tissue dependence of the absorbed dose has been investigated with vacuum phantom for breast, compact bone, blood, lung, thyroid, soft tissue, testis, and muscle. No significant variation is noted at 5 cm of radial distance in this regard while comparing the two sources except for lung tissues. The true dose rates are calculated with considering photon as well as electron transport using
Optimization of the Monte Carlo code for modeling of photon migration in tissue.
Zołek, Norbert S; Liebert, Adam; Maniewski, Roman
2006-10-01
The Monte Carlo method is frequently used to simulate light transport in turbid media because of its simplicity and flexibility, allowing to analyze complicated geometrical structures. Monte Carlo simulations are, however, time consuming because of the necessity to track the paths of individual photons. The time consuming computation is mainly associated with the calculation of the logarithmic and trigonometric functions as well as the generation of pseudo-random numbers. In this paper, the Monte Carlo algorithm was developed and optimized, by approximation of the logarithmic and trigonometric functions. The approximations were based on polynomial and rational functions, and the errors of these approximations are less than 1% of the values of the original functions. The proposed algorithm was verified by simulations of the time-resolved reflectance at several source-detector separations. The results of the calculation using the approximated algorithm were compared with those of the Monte Carlo simulations obtained with an exact computation of the logarithm and trigonometric functions as well as with the solution of the diffusion equation. The errors of the moments of the simulated distributions of times of flight of photons (total number of photons, mean time of flight and variance) are less than 2% for a range of optical properties, typical of living tissues. The proposed approximated algorithm allows to speed up the Monte Carlo simulations by a factor of 4. The developed code can be used on parallel machines, allowing for further acceleration.
Development of a Fully-Automated Monte Carlo Burnup Code Monteburns
International Nuclear Information System (INIS)
Poston, D.I.; Trellue, H.R.
1999-01-01
Several computer codes have been developed to perform nuclear burnup calculations over the past few decades. In addition, because of advances in computer technology, it recently has become more desirable to use Monte Carlo techniques for such problems. Monte Carlo techniques generally offer two distinct advantages over discrete ordinate methods: (1) the use of continuous energy cross sections and (2) the ability to model detailed, complex, three-dimensional (3-D) geometries. These advantages allow more accurate burnup results to be obtained, provided that the user possesses the required computing power (which is required for discrete ordinate methods as well). Several linkage codes have been written that combine a Monte Carlo N-particle transport code (such as MCNP TM ) with a radioactive decay and burnup code. This paper describes one such code that was written at Los Alamos National Laboratory: monteburns. Monteburns links MCNP with the isotope generation and depletion code ORIGEN2. The basis for the development of monteburns was the need for a fully automated code that could perform accurate burnup (and other) calculations for any 3-D system (accelerator-driven or a full reactor core). Before the initial development of monteburns, a list of desired attributes was made and is given below. o The code should be fully automated (that is, after the input is set up, no further user interaction is required). . The code should allow for the irradiation of several materials concurrently (each material is evaluated collectively in MCNP and burned separately in 0RIGEN2). o The code should allow the transfer of materials (shuffling) between regions in MCNP. . The code should allow any materials to be added or removed before, during, or after each step in an automated fashion. . The code should not require the user to provide input for 0RIGEN2 and should have minimal MCNP input file requirements (other than a working MCNP deck). . The code should be relatively easy to use
The Physical Models and Statistical Procedures Used in the RACER Monte Carlo Code
International Nuclear Information System (INIS)
Sutton, T.M.; Brown, F.B.; Bischoff, F.G.; MacMillan, D.B.; Ellis, C.L.; Ward, J.T.; Ballinger, C.T.; Kelly, D.J.; Schindler, L.
1999-01-01
This report describes the MCV (Monte Carlo - Vectorized)Monte Carlo neutron transport code [Brown, 1982, 1983; Brown and Mendelson, 1984a]. MCV is a module in the RACER system of codes that is used for Monte Carlo reactor physics analysis. The MCV module contains all of the neutron transport and statistical analysis functions of the system, while other modules perform various input-related functions such as geometry description, material assignment, output edit specification, etc. MCV is very closely related to the 05R neutron Monte Carlo code [Irving et al., 1965] developed at Oak Ridge National Laboratory. 05R evolved into the 05RR module of the STEMB system, which was the forerunner of the RACER system. Much of the overall logic and physics treatment of 05RR has been retained and, indeed, the original verification of MCV was achieved through comparison with STEMB results. MCV has been designed to be very computationally efficient [Brown, 1981, Brown and Martin, 1984b; Brown, 1986]. It was originally programmed to make use of vector-computing architectures such as those of the CDC Cyber- 205 and Cray X-MP. MCV was the first full-scale production Monte Carlo code to effectively utilize vector-processing capabilities. Subsequently, MCV was modified to utilize both distributed-memory [Sutton and Brown, 1994] and shared memory parallelism. The code has been compiled and run on platforms ranging from 32-bit UNIX workstations to clusters of 64-bit vector-parallel supercomputers. The computational efficiency of the code allows the analyst to perform calculations using many more neutron histories than is practical with most other Monte Carlo codes, thereby yielding results with smaller statistical uncertainties. MCV also utilizes variance reduction techniques such as survival biasing, splitting, and rouletting to permit additional reduction in uncertainties. While a general-purpose neutron Monte Carlo code, MCV is optimized for reactor physics calculations. It has the
Improvements of MCOR: A Monte Carlo depletion code system for fuel assembly reference calculations
Energy Technology Data Exchange (ETDEWEB)
Tippayakul, C.; Ivanov, K. [Pennsylvania State Univ., Univ. Park (United States); Misu, S. [AREVA NP GmbH, An AREVA and SIEMENS Company, Erlangen (Germany)
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)
Implementation of a Monte Carlo based inverse planning model for clinical IMRT with MCNP code
International Nuclear Information System (INIS)
He, Tongming Tony
2003-01-01
Inaccurate dose calculations and limitations of optimization algorithms in inverse planning introduce systematic and convergence errors to treatment plans. This work was to implement a Monte Carlo based inverse planning model for clinical IMRT aiming to minimize the aforementioned errors. The strategy was to precalculate the dose matrices of beamlets in a Monte Carlo based method followed by the optimization of beamlet intensities. The MCNP 4B (Monte Carlo N-Particle version 4B) code was modified to implement selective particle transport and dose tallying in voxels and efficient estimation of statistical uncertainties. The resulting performance gain was over eleven thousand times. Due to concurrent calculation of multiple beamlets of individual ports, hundreds of beamlets in an IMRT plan could be calculated within a practical length of time. A finite-sized point source model provided a simple and accurate modeling of treatment beams. The dose matrix calculations were validated through measurements in phantoms. Agreements were better than 1.5% or 0.2 cm. The beamlet intensities were optimized using a parallel platform based optimization algorithm that was capable of escape from local minima and preventing premature convergence. The Monte Carlo based inverse planning model was applied to clinical cases. The feasibility and capability of Monte Carlo based inverse planning for clinical IMRT was demonstrated. Systematic errors in treatment plans of a commercial inverse planning system were assessed in comparison with the Monte Carlo based calculations. Discrepancies in tumor doses and critical structure doses were up to 12% and 17%, respectively. The clinical importance of Monte Carlo based inverse planning for IMRT was demonstrated
International Nuclear Information System (INIS)
Kashima, Takao; Suyama, Kenya; Takada, Tomoyuki
2015-03-01
There have been two versions of SWAT depending on details of its development history: the revised SWAT that uses the deterministic calculation code SRAC as a neutron transportation solver, and the SWAT3.1 that uses the continuous energy Monte Carlo code MVP or MCNP5 for the same purpose. It takes several hours, however, to execute one calculation by the continuous energy Monte Carlo code even on the super computer of the Japan Atomic Energy Agency. Moreover, two-dimensional burnup calculation is not practical using the revised SWAT because it has problems on production of effective cross section data and applying them to arbitrary fuel geometry when a calculation model has multiple burnup zones. Therefore, SWAT4.0 has been developed by adding, to SWAT3.1, a function to utilize the deterministic code SARC2006, which has shorter calculation time, as an outer module of neutron transportation solver for burnup calculation. SWAT4.0 has been enabled to execute two-dimensional burnup calculation by providing an input data template of SRAC2006 to SWAT4.0 input data, and updating atomic number densities of burnup zones in each burnup step. This report describes outline, input data instruction, and examples of calculations of SWAT4.0. (author)
Electron transport in radiotherapy using local-to-global Monte Carlo
International Nuclear Information System (INIS)
Svatos, M.M.; Chandler, W.P.; Siantar, C.L.H.; Rathkopf, J.A.; Ballinger, C.T.
1994-09-01
Local-to-Global (L-G) Monte Carlo methods are a way to make three-dimensional electron transport both fast and accurate relative to other Monte Carlo methods. This is achieved by breaking the simulation into two stages: a local calculation done over small geometries having the size and shape of the ''steps'' to be taken through the mesh; and a global calculation which relies on a stepping code that samples the stored results of the local calculation. The increase in speed results from taking fewer steps in the global calculation than required by ordinary Monte Carlo codes and by speeding up the calculation per step. The potential for accuracy comes from the ability to use long runs of detailed codes to compile probability distribution functions (PDFs) in the local calculation. Specific examples of successful Local-to-Global algorithms are given
Discrete Diffusion Monte Carlo for Electron Thermal Transport
Chenhall, Jeffrey; Cao, Duc; Wollaeger, Ryan; Moses, Gregory
2014-10-01
The iSNB (implicit Schurtz Nicolai Busquet electron thermal transport method of Cao et al. is adapted to a Discrete Diffusion Monte Carlo (DDMC) solution method for eventual inclusion in a hybrid IMC-DDMC (Implicit Monte Carlo) method. The hybrid method will combine the efficiency of a diffusion method in short mean free path regions with the accuracy of a transport method in long mean free path regions. The Monte Carlo nature of the approach allows the algorithm to be massively parallelized. Work to date on the iSNB-DDMC method will be presented. This work was supported by Sandia National Laboratory - Albuquerque.
Vectorization of phase space Monte Carlo code in FACOM vector processor VP-200
International Nuclear Information System (INIS)
Miura, Kenichi
1986-01-01
This paper describes the vectorization techniques for Monte Carlo codes in Fujitsu's Vector Processor System. The phase space Monte Carlo code FOWL is selected as a benchmark, and scalar and vector performances are compared. The vectorized kernel Monte Carlo routine which contains heavily nested IF tests runs up to 7.9 times faster in vector mode than in scalar mode. The overall performance improvement of the vectorized FOWL code over the original scalar code reaches 3.3. The results of this study strongly indicate that supercomputer can be a powerful tool for Monte Carlo simulations in high energy physics. (Auth.)
Advances in Monte-Carlo code TRIPOLI-4®'s treatment of the electromagnetic cascade
Mancusi, Davide; Bonin, Alice; Hugot, François-Xavier; Malouch, Fadhel
2018-01-01
TRIPOLI-4® is a Monte-Carlo particle-transport code developed at CEA-Saclay (France) that is employed in the domains of nuclear-reactor physics, criticality-safety, shielding/radiation protection and nuclear instrumentation. The goal of this paper is to report on current developments, validation and verification made in TRIPOLI-4 in the electron/positron/photon sector. The new capabilities and improvements concern refinements to the electron transport algorithm, the introduction of a charge-deposition score, the new thick-target bremsstrahlung option, the upgrade of the bremsstrahlung model and the improvement of electron angular straggling at low energy. The importance of each of the developments above is illustrated by comparisons with calculations performed with other codes and with experimental data.
International Nuclear Information System (INIS)
Koo, Bon Seung; Lee, Kyung Hoon; Song, Jae Seung; Park, Sang Yoon
2013-01-01
In this paper, the basic nuclear characteristics of major emitter materials were surveyed. In addition, preliminary calculations of Cobalt-Vanadium fixed incore detector were performed using the Monte Carlo code. Calculational results were cross-checked by KARMA. KARMA is a two-dimensional multigroup transport theory code developed by the KAERI and approved by Korean regularity agency to be employed as a nuclear design tool for a Korean commercial pressurizer water reactor. The nuclear characteristics of the major emitter materials were surveyed, and preliminary calculations of the hybrid fixed incore detector were performed with the MCNP code. The eigenvalue and pin-by-pin fission power distributions were calculated and showed good agreement with the KARMA calculation results. As future work, gamma power distributions as well as several types of XS of the emitter, insulator, and collector regions for a Co-V ICI assembly will be evaluated and compared
A Fano cavity test for Monte Carlo proton transport algorithms
International Nuclear Information System (INIS)
Sterpin, Edmond; Sorriaux, Jefferson; Souris, Kevin; Vynckier, Stefaan; Bouchard, Hugo
2014-01-01
Purpose: In the scope of reference dosimetry of radiotherapy beams, Monte Carlo (MC) simulations are widely used to compute ionization chamber dose response accurately. Uncertainties related to the transport algorithm can be verified performing self-consistency tests, i.e., the so-called “Fano cavity test.” The Fano cavity test is based on the Fano theorem, which states that under charged particle equilibrium conditions, the charged particle fluence is independent of the mass density of the media as long as the cross-sections are uniform. Such tests have not been performed yet for MC codes simulating proton transport. The objectives of this study are to design a new Fano cavity test for proton MC and to implement the methodology in two MC codes: Geant4 and PENELOPE extended to protons (PENH). Methods: The new Fano test is designed to evaluate the accuracy of proton transport. Virtual particles with an energy ofE 0 and a mass macroscopic cross section of (Σ)/(ρ) are transported, having the ability to generate protons with kinetic energy E 0 and to be restored after each interaction, thus providing proton equilibrium. To perform the test, the authors use a simplified simulation model and rigorously demonstrate that the computed cavity dose per incident fluence must equal (ΣE 0 )/(ρ) , as expected in classic Fano tests. The implementation of the test is performed in Geant4 and PENH. The geometry used for testing is a 10 × 10 cm 2 parallel virtual field and a cavity (2 × 2 × 0.2 cm 3 size) in a water phantom with dimensions large enough to ensure proton equilibrium. Results: For conservative user-defined simulation parameters (leading to small step sizes), both Geant4 and PENH pass the Fano cavity test within 0.1%. However, differences of 0.6% and 0.7% were observed for PENH and Geant4, respectively, using larger step sizes. For PENH, the difference is attributed to the random-hinge method that introduces an artificial energy straggling if step size is not
International Nuclear Information System (INIS)
Both, J.P.; Nimal, J.C.; Vergnaud, T.
1990-01-01
We discuss an automated biasing procedure for generating the parameters necessary to achieve efficient Monte Carlo biasing shielding calculations. The biasing techniques considered here are exponential transform and collision biasing deriving from the concept of the biased game based on the importance function. We use a simple model of the importance function with exponential attenuation as the distance to the detector increases. This importance function is generated on a three-dimensional mesh including geometry and with graph theory algorithms. This scheme is currently being implemented in the third version of the neutron and gamma ray transport code TRIPOLI-3. (author)
Monte Carlo transport of electrons and positrons through thin foils
International Nuclear Information System (INIS)
Legarda, F.; Idoeta, R.
2000-01-01
In the different measurements made with electrons traversing matter it becomes useful the knowledge of its transmission through that medium, their paths and their angular distribution through matter so as to process and get information about the traversed medium and to improve and innovate the techniques that employ electrons, as medical applications or materials irradiation. This work presents a simulation of the transport of beams of electrons and positrons through thin foils using an analog Monte Carlo code that simulates in a detailed way every electron movement or interaction in matter. As those particles penetrate thin absorbers it has been assumed that they interact with matter only through elastic scattering, with negligible energy loss. This type of interaction has been described quite precisely because its angular form influences very much the angular distribution of electrons and positrons in matter. With this code it has been calculated the number of particles, with energies between 100 and 3000 keV, that are transmitted through different media of various thicknesses as well as its angular distribution, showing a good agreement with experimental data. The discrepancies are less than 5% for thicknesses lower than about 30% of the corresponding range in the tested material. As elastic scattering is very anisotropic, angular distributions resemble a collimated incident beam for very thin foils becoming slowly more isotropic when absorber thickness is increased. (author)
Monte Carlo simulation of the turbulent transport of airborne contaminants
International Nuclear Information System (INIS)
Watson, C.W.; Barr, S.
1975-09-01
A generalized, three-dimensional Monte Carlo model and computer code (SPOOR) are described for simulating atmospheric transport and dispersal of small pollutant clouds. A cloud is represented by a large number of particles that we track by statistically sampling simulated wind and turbulence fields. These fields are based on generalized wind data for large-scale flow and turbulent energy spectra for the micro- and mesoscales. The large-scale field can be input from a climatological data base, or by means of real-time analyses, or from a separate, subjectively defined data base. We introduce the micro- and mesoscale wind fluctuations through a power spectral density, to include effects from a broad spectrum of turbulent-energy scales. The role of turbulence is simulated in both meander and dispersal. Complex flow fields and time-dependent diffusion rates are accounted for naturally, and shear effects are simulated automatically in the ensemble of particle trajectories. An important adjunct has been the development of computer-graphics displays. These include two- and three-dimensional (perspective) snapshots and color motion pictures of particle ensembles, plus running displays of differential and integral cloud characteristics. The model's versatility makes it a valuable atmospheric research tool that we can adapt easily into broader, multicomponent systems-analysis codes. Removal, transformation, dry or wet deposition, and resuspension of contaminant particles can be readily included
International Nuclear Information System (INIS)
Chetty, Indrin J.; Moran, Jean M.; Nurushev, Teamor S.; McShan, Daniel L.; Fraass, Benedick A.; Wilderman, Scott J.; Bielajew, Alex F.
2002-01-01
A comprehensive set of measurements and calculations has been conducted to investigate the accuracy of the Dose Planning Method (DPM) Monte Carlo code for electron beam dose calculations in heterogeneous media. Measurements were made using 10 MeV and 50 MeV minimally scattered, uncollimated electron beams from a racetrack microtron. Source distributions for the Monte Carlo calculations were reconstructed from in-air ion chamber scans and then benchmarked against measurements in a homogeneous water phantom. The in-air spatial distributions were found to have FWHM of 4.7 cm and 1.3 cm, at 100 cm from the source, for the 10 MeV and 50 MeV beams respectively. Energy spectra for the electron beams were determined by simulating the components of the microtron treatment head using the code MCNP4B. Profile measurements were made using an ion chamber in a water phantom with slabs of lung or bone-equivalent materials submerged at various depths. DPM calculations are, on average, within 2% agreement with measurement for all geometries except for the 50 MeV incident on a 6 cm lung-equivalent slab. Measurements using approximately monoenergetic, 50 MeV, 'pencil-beam'-type electrons in heterogeneous media provide conditions for maximum electronic disequilibrium and hence present a stringent test of the code's electron transport physics; the agreement noted between calculation and measurement illustrates that the DPM code is capable of accurate dose calculation even under such conditions. (author)
International Nuclear Information System (INIS)
Kruijf, W.J.M. de; Janssen, A.J.
1994-01-01
Very accurate Mote Carlo calculations with Monte Carlo Code have been performed to serve as reference for benchmark calculations on resonance absorption by U 238 in a typical PWR pin-cell geometry. Calculations with the energy-pointwise slowing down code calculates the resonance absorption accurately. Calculations with the multigroup discrete ordinates code XSDRN show that accurate results can only be achieved with a very fine energy mesh. (authors). 9 refs., 5 figs., 2 tabs
Monte Carlo simulation in UWB1 depletion code
International Nuclear Information System (INIS)
Lovecky, M.; Prehradny, J.; Jirickova, J.; Skoda, R.
2015-01-01
U W B 1 depletion code is being developed as a fast computational tool for the study of burnable absorbers in the University of West Bohemia in Pilsen, Czech Republic. In order to achieve higher precision, the newly developed code was extended by adding a Monte Carlo solver. Research of fuel depletion aims at development and introduction of advanced types of burnable absorbers in nuclear fuel. Burnable absorbers (BA) allow the compensation of the initial reactivity excess of nuclear fuel and result in an increase of fuel cycles lengths with higher enriched fuels. The paper describes the depletion calculations of VVER nuclear fuel doped with rare earth oxides as burnable absorber based on performed depletion calculations, rare earth oxides are divided into two equally numerous groups, suitable burnable absorbers and poisoning absorbers. According to residual poisoning and BA reactivity worth, rare earth oxides marked as suitable burnable absorbers are Nd, Sm, Eu, Gd, Dy, Ho and Er, while poisoning absorbers include Sc, La, Lu, Y, Ce, Pr and Tb. The presentation slides have been added to the article
Energy Technology Data Exchange (ETDEWEB)
Baker, Randal Scott [Univ. of Arizona, Tucson, AZ (United States)
1990-01-01
The neutron transport equation is solved by a hybrid method that iteratively couples regions where deterministic (S_{N}) and stochastic (Monte Carlo) methods are applied. Unlike previous hybrid methods, the Monte Carlo and S_{N} regions are fully coupled in the sense that no assumption is made about geometrical separation or decoupling. The hybrid method provides a new means of solving problems involving both optically thick and optically thin regions that neither Monte Carlo nor S_{N} is well suited for by themselves. The fully coupled Monte Carlo/S_{N} technique consists of defining spatial and/or energy regions of a problem in which either a Monte Carlo calculation or an S_{N} calculation is to be performed. The Monte Carlo region may comprise the entire spatial region for selected energy groups, or may consist of a rectangular area that is either completely or partially embedded in an arbitrary S_{N} region. The Monte Carlo and S_{N} regions are then connected through the common angular boundary fluxes, which are determined iteratively using the response matrix technique, and volumetric sources. The hybrid method has been implemented in the S_{N} code TWODANT by adding special-purpose Monte Carlo subroutines to calculate the response matrices and volumetric sources, and linkage subrountines to carry out the interface flux iterations. The common angular boundary fluxes are included in the S_{N} code as interior boundary sources, leaving the logic for the solution of the transport flux unchanged, while, with minor modifications, the diffusion synthetic accelerator remains effective in accelerating S_{N} calculations. The special-purpose Monte Carlo routines used are essentially analog, with few variance reduction techniques employed. However, the routines have been successfully vectorized, with approximately a factor of five increase in speed over the non-vectorized version.
SWAT2: The improved SWAT code system by incorporating the continuous energy Monte Carlo code MVP
International Nuclear Information System (INIS)
Mochizuki, Hiroki; Suyama, Kenya; Okuno, Hiroshi
2003-01-01
SWAT is a code system, which performs the burnup calculation by the combination of the neutronics calculation code, SRAC95 and the one group burnup calculation code, ORIGEN2.1. The SWAT code system can deal with the cell geometry in SRAC95. However, a precise treatment of resonance absorptions by the SRAC95 code using the ultra-fine group cross section library is not directly applicable to two- or three-dimensional geometry models, because of restrictions in SRAC95. To overcome this problem, SWAT2 which newly introduced the continuous energy Monte Carlo code, MVP into SWAT was developed. Thereby, the burnup calculation by the continuous energy in any geometry became possible. Moreover, using the 147 group cross section library called SWAT library, the reactions which are not dealt with by SRAC95 and MVP can be treated. OECD/NEA burnup credit criticality safety benchmark problems Phase-IB (PWR, a single pin cell model) and Phase-IIIB (BWR, fuel assembly model) were calculated as a verification of SWAT2, and the results were compared with the average values of calculation results of burnup calculation code of each organization. Through two benchmark problems, it was confirmed that SWAT2 was applicable to the burnup calculation of the complicated geometry. (author)
Sub-step methodology for coupled Monte Carlo depletion and thermal hydraulic codes
International Nuclear Information System (INIS)
Kotlyar, D.; Shwageraus, E.
2016-01-01
Highlights: • Discretization of time in coupled MC codes determines the results’ accuracy. • The error is due to lack of information regarding the time-dependent reaction rates. • The proposed sub-step method considerably reduces the time discretization error. • No additional MC transport solutions are required within the time step. • The reaction rates are varied as functions of nuclide densities and TH conditions. - Abstract: The governing procedure in coupled Monte Carlo (MC) codes relies on discretization of the simulation time into time steps. Typically, the MC transport solution at discrete points will generate reaction rates, which in most codes are assumed to be constant within the time step. This assumption can trigger numerical instabilities or result in a loss of accuracy, which, in turn, would require reducing the time steps size. This paper focuses on reducing the time discretization error without requiring additional MC transport solutions and hence with no major computational overhead. The sub-step method presented here accounts for the reaction rate variation due to the variation in nuclide densities and thermal hydraulic (TH) conditions. This is achieved by performing additional depletion and TH calculations within the analyzed time step. The method was implemented in BGCore code and subsequently used to analyze a series of test cases. The results indicate that computational speedup of up to a factor of 10 may be achieved over the existing coupling schemes.
MORSE-C, Neutron Transport, Gamma Transport for Criticality Calculation by Monte-Carlo Method
International Nuclear Information System (INIS)
2002-01-01
1 - Description of program or function: MORSE-C is a Monte-Carlo code to solve the multiple energy group form of the Boltzmann transport equation in order to obtain the eigenvalue (multiplication) when fissionable materials are present. Cross sections for up to 100 energy groups may be employed. The angular scattering is treated by the usual Legendre expansion as used in the discrete ordinates codes. Up-scattering may be specified. The geometry is defined by relationships to general 1. or 2. degree surfaces. Array units may be specified. Output includes, besides the usual values of input quantities, plots of the geometry, calculated volumes and masses, and graphs of results to assist the user in determining the correctness of the problem's solution
Experimental transport analysis code system in JT-60
International Nuclear Information System (INIS)
Hirayama, Toshio; Shimizu, Katsuhiro; Tani, Keiji; Shirai, Hiroshi; Kikuchi, Mitsuru
1988-03-01
Transport analysis codes have been developed in order to study confinement properties related to particle and energy balance in ohmically and neutral beam heated plasmas of JT-60. The analysis procedure is divided into three steps as follows: 1) LOOK ; The shape of the plasma boundary is identified with a fast boundary identification code of FBI by using magnetic data, and flux surfaces are calculated with a MHD equilibrium code of SELENE. The diagnostic data are mapped to flux surfaces for neutral beam heating calculation and/or for radial transport analysis. 2) OFMC ; On the basis of transformed data, an orbit following Monte Carlo code of OFMC calculates both profiles of power deposition and particle source of neutral beam injected into a plasma. 3) SCOOP ; In the last stage, a one dimensional transport code of SCOOP solves particle and energy balance for electron and ion, in order to evaluate transport coefficients as well as global parameters such as energy confinement time and the stored energy. The analysis results are provided to a data bank of DARTS that is used to find an overview of important consideration on confinement with a regression analysis code of RAC. (author)
International Nuclear Information System (INIS)
Macdonald, J.L.
1975-08-01
Statistical and deterministic pattern recognition systems are designed to classify the state space of a Monte Carlo transport problem into importance regions. The surfaces separating the regions can be used for particle splitting and Russian roulette in state space in order to reduce the variance of the Monte Carlo tally. Computer experiments are performed to evaluate the performance of the technique using one and two dimensional Monte Carlo problems. Additional experiments are performed to determine the sensitivity of the technique to various pattern recognition and Monte Carlo problem dependent parameters. A system for applying the technique to a general purpose Monte Carlo code is described. An estimate of the computer time required by the technique is made in order to determine its effectiveness as a variance reduction device. It is recommended that the technique be further investigated in a general purpose Monte Carlo code. (auth)
Review of the Monte Carlo and deterministic codes in radiation protection and dosimetry
Energy Technology Data Exchange (ETDEWEB)
Tagziria, H
2000-02-01
Modelling a physical system can be carried out either stochastically or deterministically. An example of the former method is the Monte Carlo technique, in which statistically approximate methods are applied to exact models. No transport equation is solved as individual particles are simulated and some specific aspect (tally) of their average behaviour is recorded. The average behaviour of the physical system is then inferred using the central limit theorem. In contrast, deterministic codes use mathematically exact methods that are applied to approximate models to solve the transport equation for the average particle behaviour. The physical system is subdivided in boxes in the phase-space system and particles are followed from one box to the next. The smaller the boxes the better the approximations become. Although the Monte Carlo method has been used for centuries, its more recent manifestation has really emerged from the Manhattan project of the Word War II. Its invention is thought to be mainly due to Metropolis, Ulah (through his interest in poker), Fermi, von Neuman andRichtmeyer. Over the last 20 years or so, the Monte Carlo technique has become a powerful tool in radiation transport. This is due to users taking full advantage of richer cross section data, more powerful computers and Monte Carlo techniques for radiation transport, with high quality physics and better known source spectra. This method is a common sense approach to radiation transport and its success and popularity is quite often also due to necessity, because measurements are not always possible or affordable. In the Monte Carlo method, which is inherently realistic because nature is statistical, a more detailed physics is made possible by isolation of events while rather elaborate geometries can be modelled. Provided that the physics is correct, a simulation is exactly analogous to an experimenter counting particles. In contrast to the deterministic approach, however, a disadvantage of the
SU-E-T-323: The FLUKA Monte Carlo Code in Ion Beam Therapy
Energy Technology Data Exchange (ETDEWEB)
Rinaldi, I [Heidelberg University Hospital (Germany); Ludwig-Maximilian University Munich (Germany)
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
Development of particle and heavy ion transport code system
International Nuclear Information System (INIS)
Niita, Koji
2004-01-01
Particle and heavy ion transport code system (PHITS) is 3 dimension general purpose Monte Carlo simulation codes for description of transport and reaction of particle and heavy ion in materials. It is developed on the basis of NMTC/JAM for design and safety of J-PARC. What is PHITS, it's physical process, physical models and development process of PHITC code are described. For examples of application, evaluation of neutron optics, cancer treatment by heavy particle ray and cosmic radiation are stated. JAM and JQMD model are used as the physical model. Neutron motion in six polar magnetic field and gravitational field, PHITC simulation of trace of C 12 beam and secondary neutron track of small model of cancer treatment device in HIMAC and neutron flux in Space Shuttle are explained. (S.Y.)
Energy Technology Data Exchange (ETDEWEB)
Belicev, P [Vojnotehnicki Inst., Belgrade (Yugoslavia)
1988-07-01
An outline of the problems encountered in the multigroup calculations of the neutron transport in the resonance region is given. The difference between subgroup and multigroup approximation is described briefly. The features of the Monte Carlo code SUBGR are presented. The results of the calculations of the neutron transmission and albedo for infinite iron slabs are given. (author)
International Nuclear Information System (INIS)
Hoogenboom, J. Eduard
2003-01-01
Adjoint Monte Carlo may be a useful alternative to regular Monte Carlo calculations in cases where a small detector inhibits an efficient Monte Carlo calculation as only very few particle histories will cross the detector. However, in general purpose Monte Carlo codes, normally only the multigroup form of adjoint Monte Carlo is implemented. In this article the general methodology for continuous-energy adjoint Monte Carlo neutron transport is reviewed and extended for photon and coupled neutron-photon transport. In the latter cases the discrete photons generated by annihilation or by neutron capture or inelastic scattering prevent a direct application of the general methodology. Two successive reaction events must be combined in the selection process to accommodate the adjoint analog of a reaction resulting in a photon with a discrete energy. Numerical examples illustrate the application of the theory for some simplified problems
Monte Carlo Studies of Electron Transport In Semiconductor Nanostructures
Tierney, Brian David
An Ensemble Monte Carlo (EMC) computer code has been developed to simulate, semi-classically, spin-dependent electron transport in quasi two-dimensional (2D) III-V semiconductors. The code accounts for both three-dimensional (3D) and quasi-2D transport, utilizing either 3D or 2D scattering mechanisms, as appropriate. Phonon, alloy, interface roughness, and impurity scattering mechanisms are included, accounting for the Pauli Exclusion Principle via a rejection algorithm. The 2D carrier states are calculated via a self-consistent 1D Schrodinger-3D-Poisson solution in which the charge distribution of the 2D carriers in the quantization direction is taken as the spatial distribution of the squared envelope functions within the Hartree approximation. The wavefunctions, subband energies, and 2D scattering rates are updated periodically by solving a series of 1D Schrodinger wave equations (SWE) over the real-space domain of the device at fixed time intervals. The electrostatic potential is updated by periodically solving the 3D Poisson equation. Spin-polarized transport is modeled via a spin density-matrix formalism that accounts for D'yakanov-Perel (DP) scattering. Also, the code allows for the easy inclusion of additional scattering mechanisms and structural modifications to devices. As an application of the simulator, the current voltage characteristics of an InGaAs/InAlAs HEMT are simulated, corresponding to nanoscale III-V HEMTs currently being fabricated by Intel Corporation. The comparative effects of various scattering parameters, material properties and structural attributes are investigated and compared with experiments where reasonable agreement is obtained. The spatial evolution of spin-polarized carriers in prototypical Spin Field Effect Transistor (SpinFET) devices is then simulated. Studies of the spin coherence times in quasi-2D structures is first investigated and compared to experimental results. It is found that the simulated spin coherence times for
A flexible coupling scheme for Monte Carlo and thermal-hydraulics codes
Energy Technology Data Exchange (ETDEWEB)
Hoogenboom, J. Eduard, E-mail: J.E.Hoogenboom@tudelft.nl [Delft University of Technology (Netherlands); Ivanov, Aleksandar; Sanchez, Victor, E-mail: Aleksandar.Ivanov@kit.edu, E-mail: Victor.Sanchez@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Eggenstein-Leopoldshafen (Germany); Diop, Cheikh, E-mail: Cheikh.Diop@cea.fr [CEA/DEN/DANS/DM2S/SERMA, Commissariat a l' Energie Atomique, Gif-sur-Yvette (France)
2011-07-01
A coupling scheme between a Monte Carlo code and a thermal-hydraulics code is being developed within the European NURISP project for comprehensive and validated reactor analysis. The scheme is flexible as it allows different Monte Carlo codes and different thermal-hydraulics codes to be used. At present the MCNP and TRIPOLI4 Monte Carlo codes can be used and the FLICA4 and SubChanFlow thermal-hydraulics codes. For all these codes only an original executable is necessary. A Python script drives the iterations between Monte Carlo and thermal-hydraulics calculations. It also calls a conversion program to merge a master input file for the Monte Carlo code with the appropriate temperature and coolant density data from the thermal-hydraulics calculation. Likewise it calls another conversion program to merge a master input file for the thermal-hydraulics code with the power distribution data from the Monte Carlo calculation. Special attention is given to the neutron cross section data for the various required temperatures in the Monte Carlo calculation. Results are shown for an infinite lattice of PWR fuel pin cells and a 3 x 3 fuel BWR pin cell cluster. Various possibilities for further improvement and optimization of the coupling system are discussed. (author)
A flexible coupling scheme for Monte Carlo and thermal-hydraulics codes
International Nuclear Information System (INIS)
Hoogenboom, J. Eduard; Ivanov, Aleksandar; Sanchez, Victor; Diop, Cheikh
2011-01-01
A coupling scheme between a Monte Carlo code and a thermal-hydraulics code is being developed within the European NURISP project for comprehensive and validated reactor analysis. The scheme is flexible as it allows different Monte Carlo codes and different thermal-hydraulics codes to be used. At present the MCNP and TRIPOLI4 Monte Carlo codes can be used and the FLICA4 and SubChanFlow thermal-hydraulics codes. For all these codes only an original executable is necessary. A Python script drives the iterations between Monte Carlo and thermal-hydraulics calculations. It also calls a conversion program to merge a master input file for the Monte Carlo code with the appropriate temperature and coolant density data from the thermal-hydraulics calculation. Likewise it calls another conversion program to merge a master input file for the thermal-hydraulics code with the power distribution data from the Monte Carlo calculation. Special attention is given to the neutron cross section data for the various required temperatures in the Monte Carlo calculation. Results are shown for an infinite lattice of PWR fuel pin cells and a 3 x 3 fuel BWR pin cell cluster. Various possibilities for further improvement and optimization of the coupling system are discussed. (author)
OpenMC: A state-of-the-art Monte Carlo code for research and development
International Nuclear Information System (INIS)
Romano, Paul K.; Horelik, Nicholas E.; Herman, Bryan R.; Nelson, Adam G.; Forget, Benoit; Smith, Kord
2015-01-01
Highlights: • OpenMC is an open source Monte Carlo particle transport code. • Solid geometry and continuous-energy physics allow high-fidelity simulations. • Development has focused on high performance and modern I/O techniques. • OpenMC is capable of scaling up to hundreds of thousands of processors. • Other features include plotting, CMFD acceleration, and variance reduction. - Abstract: 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
Energy Technology Data Exchange (ETDEWEB)
Moskvin, V; Tsiamas, P; Axente, M; Farr, J [St. Jude Children’s Research Hospital, Memphis, TN (United States); Stewart, R [University of Washington, Seattle, WA. (United States)
2015-06-15
Purpose: One of the more critical initiating events for reproductive cell death is the creation of a DNA double strand break (DSB). In this study, we present a computationally efficient way to determine spatial variations in the relative biological effectiveness (RBE) of proton therapy beams within the FLUKA Monte Carlo (MC) code. Methods: We used the independently tested Monte Carlo Damage Simulation (MCDS) developed by Stewart and colleagues (Radiat. Res. 176, 587–602 2011) to estimate the RBE for DSB induction of monoenergetic protons, tritium, deuterium, hellium-3, hellium-4 ions and delta-electrons. The dose-weighted (RBE) coefficients were incorporated into FLUKA to determine the equivalent {sup 6}°60Co γ-ray dose for representative proton beams incident on cells in an aerobic and anoxic environment. Results: We found that the proton beam RBE for DSB induction at the tip of the Bragg peak, including primary and secondary particles, is close to 1.2. Furthermore, the RBE increases laterally to the beam axis at the area of Bragg peak. At the distal edge, the RBE is in the range from 1.3–1.4 for cells irradiated under aerobic conditions and may be as large as 1.5–1.8 for cells irradiated under anoxic conditions. Across the plateau region, the recorded RBE for DSB induction is 1.02 for aerobic cells and 1.05 for cells irradiated under anoxic conditions. The contribution to total effective dose from secondary heavy ions decreases with depth and is higher at shallow depths (e.g., at the surface of the skin). Conclusion: Multiscale simulation of the RBE for DSB induction provides useful insights into spatial variations in proton RBE within pristine Bragg peaks. This methodology is potentially useful for the biological optimization of proton therapy for the treatment of cancer. The study highlights the need to incorporate spatial variations in proton RBE into proton therapy treatment plans.
Energy Technology Data Exchange (ETDEWEB)
Zychor, I. [Soltan Inst. for Nuclear Studies, Otwock-Swierk (Poland)
1994-12-31
The application of a Monte Carlo method to study a transport in matter of electron and photon beams is presented, especially for electrons with energies up to 18 MeV. The SHOWME Monte Carlo code, a modified version of GEANT3 code, was used on the CONVEX C3210 computer at Swierk. It was assumed that an electron beam is mono directional and monoenergetic. Arbitrary user-defined, complex geometries made of any element or material can be used in calculation. All principal phenomena occurring when electron beam penetrates the matter are taken into account. The use of calculation for a therapeutic electron beam collimation is presented. (author). 20 refs, 29 figs.
Monte Carlo method for neutron transport calculations in graphics processing units (GPUs)
International Nuclear Information System (INIS)
Pellegrino, Esteban
2011-01-01
Monte Carlo simulation is well suited for solving the Boltzmann neutron transport equation in an inhomogeneous media for complicated geometries. However, routine applications require the computation time to be reduced to hours and even minutes in a desktop PC. The interest in adopting Graphics Processing Units (GPUs) for Monte Carlo acceleration is rapidly growing. This is due to the massive parallelism provided by the latest GPU technologies which is the most promising solution to the challenge of performing full-size reactor core analysis on a routine basis. In this study, Monte Carlo codes for a fixed-source neutron transport problem were developed for GPU environments in order to evaluate issues associated with computational speedup using GPUs. Results obtained in this work suggest that a speedup of several orders of magnitude is possible using the state-of-the-art GPU technologies. (author) [es
In-facility transport code review
International Nuclear Information System (INIS)
Spore, J.W.; Boyack, B.E.; Bohl, W.R.
1996-07-01
The following computer codes were reviewed by the In-Facility Transport Working Group for application to the in-facility transport of radioactive aerosols, flammable gases, and/or toxic gases: (1) CONTAIN, (2) FIRAC, (3) GASFLOW, (4) KBERT, and (5) MELCOR. Based on the review criteria as described in this report and the versions of each code available at the time of the review, MELCOR is the best code for the analysis of in-facility transport when multidimensional effects are not significant. When multi-dimensional effects are significant, GASFLOW should be used
International Nuclear Information System (INIS)
Androsenko, A.A.; Androsenko, P.A.; Kagalenko, I.Eh.; Mironovich, Yu.N.
1992-01-01
Consideration is given of a technique and algorithms of constructing neutron trajectories in the Monte-Carlo method taking into account the data on adjoint transport equation solution. When simulating the transport part of transfer kernel the use is made of piecewise-linear approximation of free path length density along the particle motion direction. The approach has been implemented in programs within the framework of the BRAND code system. The importance is calculated in the multigroup P 1 -approximation within the framework of the DD-30 code system. The efficiency of the developed computation technique is demonstrated by means of solution of two model problems. 4 refs.; 2 tabs
Benchmark of neutron production cross sections with Monte Carlo codes
Tsai, Pi-En; Lai, Bo-Lun; Heilbronn, Lawrence H.; Sheu, Rong-Jiun
2018-02-01
Aiming to provide critical information in the fields of heavy ion therapy, radiation shielding in space, and facility design for heavy-ion research accelerators, the physics models in three Monte Carlo simulation codes - PHITS, FLUKA, and MCNP6, were systematically benchmarked with comparisons to fifteen sets of experimental data for neutron production cross sections, which include various combinations of 12C, 20Ne, 40Ar, 84Kr and 132Xe projectiles and natLi, natC, natAl, natCu, and natPb target nuclides at incident energies between 135 MeV/nucleon and 600 MeV/nucleon. For neutron energies above 60% of the specific projectile energy per nucleon, the LAQGMS03.03 in MCNP6, the JQMD/JQMD-2.0 in PHITS, and the RQMD-2.4 in FLUKA all show a better agreement with data in heavy-projectile systems than with light-projectile systems, suggesting that the collective properties of projectile nuclei and nucleon interactions in the nucleus should be considered for light projectiles. For intermediate-energy neutrons whose energies are below the 60% projectile energy per nucleon and above 20 MeV, FLUKA is likely to overestimate the secondary neutron production, while MCNP6 tends towards underestimation. PHITS with JQMD shows a mild tendency for underestimation, but the JQMD-2.0 model with a modified physics description for central collisions generally improves the agreement between data and calculations. For low-energy neutrons (below 20 MeV), which are dominated by the evaporation mechanism, PHITS (which uses GEM linked with JQMD and JQMD-2.0) and FLUKA both tend to overestimate the production cross section, whereas MCNP6 tends to underestimate more systems than to overestimate. For total neutron production cross sections, the trends of the benchmark results over the entire energy range are similar to the trends seen in the dominate energy region. Also, the comparison of GEM coupled with either JQMD or JQMD-2.0 in the PHITS code indicates that the model used to describe the first
Monte Carlo simulation of neutron transport phenomena
International Nuclear Information System (INIS)
Srinivasan, P.
2009-01-01
Neutron transport is one of the central problems in nuclear reactor related studies and other applied sciences. Some of the major applications of neutron transport include nuclear reactor design and safety, criticality safety of fissile material handling, neutron detector design and development, nuclear medicine, assessment of radiation damage to materials, nuclear well logging, forensic analysis etc. Most reactor and dosimetry studies assume that neutrons diffuse from regions of high to low density just like gaseous molecules diffuse to regions of low concentration or heat flow from high to low temperature regions. However while treatment of gaseous or heat diffusion is quite accurately modeled, treatment of neutron transport as simple diffusion is quite limited. In simple diffusion, the neutron trajectories are irregular, random and zigzag - where as in neutron transport low reaction cross sections (1 barn- 10 -24 cm 2 ) lead to long mean free paths which again depend on the nature and irregularities of the medium. Hence a more accurate representation of the neutron transport evolved based on the Boltzmann equation of kinetic gas theory. In fact the neutron transport equation is a linearized version of the Boltzmann gas equation based on neutron conservation with seven independent variables. The transport equation is difficult to solve except in simple cases amenable to numerical methods. The diffusion (equation) approximation follows from removing the angular dependence of the neutron flux
International Nuclear Information System (INIS)
Odano, N.; Miura, T.; Yamaji, A.
1996-01-01
Measurement of activation reaction rates was carried out for fast neutrons penetrating through graphite and water from the core of JRR-4 research reactor of JAERI, with paying attention to the energy above 10 MeV. Analysis of the experiment was made using a vectorized continuous energy Monte Carlo code MVP to verify the code. The analysis shows good agreements between the measurement and calculation and the MVP code has been confirmed its validity for the fast neutron transport calculations above 10 MeV in fission neutron field. (author)
A User's Manual for MASH V1.5 - A Monte Carlo Adjoint Shielding Code System
Energy Technology Data Exchange (ETDEWEB)
C. O. Slater; J. M. Barnes; J. O. Johnson; J.D. Drischler
1998-10-01
The Monte Carlo ~djoint ~ielding Code System, MASH, calculates neutron and gamma- ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air- over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system includes the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. The current version, MASH v 1.5, is the successor to the original MASH v 1.0 code system initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the "dose importance" of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response as a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem.
Energy Technology Data Exchange (ETDEWEB)
Bordy, J M; Kodeli, I; Menard, St; Bouchet, J L; Renard, F; Martin, E; Blazy, L; Voros, S; Bochud, F; Laedermann, J P; Beaugelin, K; Makovicka, L; Quiot, A; Vermeersch, F; Roche, H; Perrin, M C; Laye, F; Bardies, M; Struelens, L; Vanhavere, F; Gschwind, R; Fernandez, F; Quesne, B; Fritsch, P; Lamart, St; Crovisier, Ph; Leservot, A; Antoni, R; Huet, Ch; Thiam, Ch; Donadille, L; Monfort, M; Diop, Ch; Ricard, M
2006-07-01
The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations.
International Nuclear Information System (INIS)
Ding, Aiping; Liu, Tianyu; Liang, Chao; Ji, Wei; Shephard, Mark S.; Xu, X George; Brown, Forrest B.
2011-01-01
Monte Carlo simulation is ideally suited for solving Boltzmann neutron transport equation in inhomogeneous media. However, routine applications require the computation time to be reduced to hours and even minutes in a desktop system. The interest in adopting GPUs for Monte Carlo acceleration is rapidly mounting, fueled partially by the parallelism afforded by the latest GPU technologies and the challenge to perform full-size reactor core analysis on a routine basis. In this study, Monte Carlo codes for a fixed-source neutron transport problem and an eigenvalue/criticality problem were developed for CPU and GPU environments, respectively, to evaluate issues associated with computational speedup afforded by the use of GPUs. The results suggest that a speedup factor of 30 in Monte Carlo radiation transport of neutrons is within reach using the state-of-the-art GPU technologies. However, for the eigenvalue/criticality problem, the speedup was 8.5. In comparison, for a task of voxelizing unstructured mesh geometry that is more parallel in nature, the speedup of 45 was obtained. It was observed that, to date, most attempts to adopt GPUs for Monte Carlo acceleration were based on naïve implementations and have not yielded the level of anticipated gains. Successful implementation of Monte Carlo schemes for GPUs will likely require the development of an entirely new code. Given the prediction that future-generation GPU products will likely bring exponentially improved computing power and performances, innovative hardware and software solutions may make it possible to achieve full-core Monte Carlo calculation within one hour using a desktop computer system in a few years. (author)
A Monte Carlo Green's function method for three-dimensional neutron transport
International Nuclear Information System (INIS)
Gamino, R.G.; Brown, F.B.; Mendelson, M.R.
1992-01-01
This paper describes a Monte Carlo transport kernel capability, which has recently been incorporated into the RACER continuous-energy Monte Carlo code. The kernels represent a Green's function method for neutron transport from a fixed-source volume out to a particular volume of interest. This method is very powerful transport technique. Also, since kernels are evaluated numerically by Monte Carlo, the problem geometry can be arbitrarily complex, yet exact. This method is intended for problems where an ex-core neutron response must be determined for a variety of reactor conditions. Two examples are ex-core neutron detector response and vessel critical weld fast flux. The response is expressed in terms of neutron transport kernels weighted by a core fission source distribution. In these types of calculations, the response must be computed for hundreds of source distributions, but the kernels only need to be calculated once. The advance described in this paper is that the kernels are generated with a highly accurate three-dimensional Monte Carlo transport calculation instead of an approximate method such as line-of-sight attenuation theory or a synthesized three-dimensional discrete ordinates solution
Development of burnup methods and capabilities in Monte Carlo code RMC
International Nuclear Information System (INIS)
She, Ding; Liu, Yuxuan; Wang, Kan; Yu, Ganglin; Forget, Benoit; Romano, Paul K.; Smith, Kord
2013-01-01
Highlights: ► The RMC code has been developed aiming at large-scale burnup calculations. ► Matrix exponential methods are employed to solve the depletion equations. ► The Energy-Bin method reduces the time expense of treating ACE libraries. ► The Cell-Mapping method is efficient to handle massive amounts of tally cells. ► Parallelized depletion is necessary for massive amounts of burnup regions. -- Abstract: The Monte Carlo burnup calculation has always been a challenging problem because of its large time consumption when applied to full-scale assembly or core calculations, and thus its application in routine analysis is limited. Most existing MC burnup codes are usually external wrappers between a MC code, e.g. MCNP, and a depletion code, e.g. ORIGEN. The code RMC is a newly developed MC code with an embedded depletion module aimed at performing burnup calculations of large-scale problems with high efficiency. Several measures have been taken to strengthen the burnup capabilities of RMC. Firstly, an accurate and efficient depletion module called DEPTH has been developed and built in, which employs the rational approximation and polynomial approximation methods. Secondly, the Energy-Bin method and the Cell-Mapping method are implemented to speed up the transport calculations with large numbers of nuclides and tally cells. Thirdly, the batch tally method and the parallelized depletion module have been utilized to better handle cases with massive amounts of burnup regions in parallel calculations. Burnup cases including a PWR pin and a 5 × 5 assembly group are calculated, thereby demonstrating the burnup capabilities of the RMC code. In addition, the computational time and memory requirements of RMC are compared with other MC burnup codes.
COG10, Multiparticle Monte Carlo Code System for Shielding and Criticality Use
International Nuclear Information System (INIS)
2007-01-01
1 - Description of program or function: COG is a modern, full-featured Monte Carlo radiation transport code which provides accurate answers to complex shielding, criticality, and activation problems. COG was written to be state-of-the-art and free of physics approximations and compromises found in earlier codes. COG is fully 3-D, uses point-wise cross sections and exact angular scattering, and allows a full range of biasing options to speed up solutions for deep penetration problems. Additionally, a criticality option is available for computing Keff for assemblies of fissile materials. ENDL or ENDFB cross section libraries may be used. COG home page: http://www-phys.llnl.gov/N_Div/COG/. Cross section libraries are included in the package. COG can use either the LLNL ENDL-90 cross section set or the ENDFB/VI set. Analytic surfaces are used to describe geometric boundaries. Parts (volumes) are described by a method of Constructive Solid Geometry. Surface types include surfaces of up to fourth order, and pseudo-surfaces such as boxes, finite cylinders, and figures of revolution. Repeated assemblies need be defined only once. Parts are visualized in cross-section and perspective picture views. Source and random-walk biasing techniques may be selected to improve solution statistics. These include source angular biasing, importance weighting, particle splitting and Russian roulette, path-length stretching, point detectors, scattered direction biasing, and forced collisions. Criticality - For a fissioning system, COG will compute Keff by transporting batches of neutrons through the system. Activation - COG can compute gamma-ray doses due to neutron-activated materials, starting with just a neutron source. Coupled Problems - COG can solve coupled problems involving neutrons, photons, and electrons. 2 - Methods:COG uses Monte Carlo methods to solve the Boltzmann transport equation for particles traveling through arbitrary 3-dimensional geometries. Neutrons, photons
Confidence interval procedures for Monte Carlo transport simulations
International Nuclear Information System (INIS)
Pederson, S.P.
1997-01-01
The problem of obtaining valid confidence intervals based on estimates from sampled distributions using Monte Carlo particle transport simulation codes such as MCNP is examined. Such intervals can cover the true parameter of interest at a lower than nominal rate if the sampled distribution is extremely right-skewed by large tallies. Modifications to the standard theory of confidence intervals are discussed and compared with some existing heuristics, including batched means normality tests. Two new types of diagnostics are introduced to assess whether the conditions of central limit theorem-type results are satisfied: the relative variance of the variance determines whether the sample size is sufficiently large, and estimators of the slope of the right tail of the distribution are used to indicate the number of moments that exist. A simulation study is conducted to quantify the relationship between various diagnostics and coverage rates and to find sample-based quantities useful in indicating when intervals are expected to be valid. Simulated tally distributions are chosen to emulate behavior seen in difficult particle transport problems. Measures of variation in the sample variance s 2 are found to be much more effective than existing methods in predicting when coverage will be near nominal rates. Batched means tests are found to be overly conservative in this regard. A simple but pathological MCNP problem is presented as an example of false convergence using existing heuristics. The new methods readily detect the false convergence and show that the results of the problem, which are a factor of 4 too small, should not be used. Recommendations are made for applying these techniques in practice, using the statistical output currently produced by MCNP
Modular, object-oriented redesign of a large-scale Monte Carlo neutron transport program
International Nuclear Information System (INIS)
Moskowitz, B.S.
2000-01-01
This paper describes the modular, object-oriented redesign of a large-scale Monte Carlo neutron transport program. This effort represents a complete 'white sheet of paper' rewrite of the code. In this paper, the motivation driving this project, the design objectives for the new version of the program, and the design choices and their consequences will be discussed. The design itself will also be described, including the important subsystems as well as the key classes within those subsystems
A Monte Carlo method using octree structure in photon and electron transport
International Nuclear Information System (INIS)
Ogawa, K.; Maeda, S.
1995-01-01
Most of the early Monte Carlo calculations in medical physics were used to calculate absorbed dose distributions, and detector responses and efficiencies. Recently, data acquisition in Single Photon Emission CT (SPECT) has been simulated by a Monte Carlo method to evaluate scatter photons generated in a human body and a collimator. Monte Carlo simulations in SPECT data acquisition are generally based on the transport of photons only because the photons being simulated are low energy, and therefore the bremsstrahlung productions by the electrons generated are negligible. Since the transport calculation of photons without electrons is much simpler than that with electrons, it is possible to accomplish the high-speed simulation in a simple object with one medium. Here, object description is important in performing the photon and/or electron transport using a Monte Carlo method efficiently. The authors propose a new description method using an octree representation of an object. Thus even if the boundaries of each medium are represented accurately, high-speed calculation of photon transport can be accomplished because the number of voxels is much fewer than that of the voxel-based approach which represents an object by a union of the voxels of the same size. This Monte Carlo code using the octree representation of an object first establishes the simulation geometry by reading octree string, which is produced by forming an octree structure from a set of serial sections for the object before the simulation; then it transports photons in the geometry. Using the code, if the user just prepares a set of serial sections for the object in which he or she wants to simulate photon trajectories, he or she can perform the simulation automatically using the suboptimal geometry simplified by the octree representation without forming the optimal geometry by handwriting
International Nuclear Information System (INIS)
Cramer, S.N.
1985-09-01
An overview of the RSIC-distributed version of the MCNP code (a soupled Monte Carlo neutron-photon code) is presented. All general features of the code, from machine hardware requirements to theoretical details, are discussed. The current nuclide cross-section and other libraries available in the standard code package are specified, and a realistic example of the flexible geometry input is given. Standard and nonstandard source, estimator, and variance-reduction procedures are outlined. Examples of correct usage and possible misuse of certain code features are presented graphically and in standard output listings. Finally, itemized summaries of sample problems, various MCNP code documentation, and future work are given
Implementation of the probability table method in a continuous-energy Monte Carlo code system
International Nuclear Information System (INIS)
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
Particle and heavy ion transport code system; PHITS
International Nuclear Information System (INIS)
Niita, Koji
2004-01-01
Intermediate and high energy nuclear data are strongly required in design study of many facilities such as accelerator-driven systems, intense pulse spallation neutron sources, and also in medical and space technology. There is, however, few evaluated nuclear data of intermediate and high energy nuclear reactions. Therefore, we have to use some models or systematics for the cross sections, which are essential ingredients of high energy particle and heavy ion transport code to estimate neutron yield, heat deposition and many other quantities of the transport phenomena in materials. We have developed general purpose particle and heavy ion transport Monte Carlo code system, PHITS (Particle and Heavy Ion Transport code System), based on the NMTC/JAM code by the collaboration of Tohoku University, JAERI and RIST. The PHITS has three important ingredients which enable us to calculate (1) high energy nuclear reactions up to 200 GeV, (2) heavy ion collision and its transport in material, (3) low energy neutron transport based on the evaluated nuclear data. In the PHITS, the cross sections of high energy nuclear reactions are obtained by JAM model. JAM (Jet AA Microscopic Transport Model) is a hadronic cascade model, which explicitly treats all established hadronic states including resonances and all hadron-hadron cross sections parametrized based on the resonance model and string model by fitting the available experimental data. The PHITS can describe the transport of heavy ions and their collisions by making use of JQMD and SPAR code. The JQMD (JAERI Quantum Molecular Dynamics) is a simulation code for nucleus nucleus collisions based on the molecular dynamics. The SPAR code is widely used to calculate the stopping powers and ranges for charged particles and heavy ions. The PHITS has included some part of MCNP4C code, by which the transport of low energy neutron, photon and electron based on the evaluated nuclear data can be described. Furthermore, the high energy nuclear
Improvement of Monte Carlo code A3MCNP for large-scale shielding problems
International Nuclear Information System (INIS)
Miyake, Y.; Ohmura, M.; Hasegawa, T.; Ueki, K.; Sato, O.; Haghighat, A.; Sjoden, G.E.
2004-01-01
A 3 MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic 'importance' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3 MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A 3 MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3 MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3 MCNP (referred to as A 3 MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3 MCNPV for a concrete cask streaming problem and a PWR dosimetry problem. (author)
Performance of the improved version of Monte Carlo Code A3MCNP for cask shielding design
International Nuclear Information System (INIS)
Hasegawa, T.; Ueki, K.; Sato, O.; Sjoden, G.E.; Miyake, Y.; Ohmura, M.; Haghighat, A.
2004-01-01
A 3 MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic ''importance'' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3 MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3 MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3 MCNP (referred to as A 3 MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3 MCNPV for cask neutron and gamma-ray shielding problem
The MCU-RFFI Monte Carlo code for reactor design applications
International Nuclear Information System (INIS)
Gomin, E.A.; Maiorov, L.V.
1995-01-01
MCU-RFFI is a general-purpose, continuous-energy, general geometry Monte Carlo code for solving external source or criticality problems for neutron transport in the energy range of 20 MeV to 10 -5 eV. The main fields of MCU-RFFI applications are as follows: (a) nuclear data validation; (b) design calculations (space reactors and other); (c) verification of design codes. MCU-RFFI is also supplied with tools to check the accuracy of design codes. These tools permit the user to calculate: the few group parameters of reactor cells, including the diffusion coefficients defined in a variety of ways, reaction rates for various nuclei, energy and space bins, and the kinetic parameters of systems, taking into account delayed neutrons. Boundary conditions include vacuum, white and specular reflection, and the condition of translational symmetry. The criticals with the neutron leakage given by the buckling vector may be calculated by solving Benoist's problem. The curve of criticality coefficient dependence on buckling may be determined during the single code run and critical buckling may be determined. Double heterogeneous systems with fuel elements containing many thousands of spherical microcells can be solved
Modified Monte Carlo procedure for particle transport problems
International Nuclear Information System (INIS)
Matthes, W.
1978-01-01
The simulation of photon transport in the atmosphere with the Monte Carlo method forms part of the EURASEP-programme. The specifications for the problems posed for a solution were such, that the direct application of the analogue Monte Carlo method was not feasible. For this reason the standard Monte Carlo procedure was modified in the sense that additional properly weighted branchings at each collision and transport process in a photon history were introduced. This modified Monte Carlo procedure leads to a clear and logical separation of the essential parts of a problem and offers a large flexibility for variance reducing techniques. More complex problems, as foreseen in the EURASEP-programme (e.g. clouds in the atmosphere, rough ocean-surface and chlorophyl-distribution in the ocean) can be handled by recoding some subroutines. This collision- and transport-splitting procedure can of course be performed differently in different space- and energy regions. It is applied here only for a homogeneous problem
Overview and applications of the Monte Carlo radiation transport kit at LLNL
International Nuclear Information System (INIS)
Sale, K. E.
1999-01-01
Modern Monte Carlo radiation transport codes can be applied to model most applications of radiation, from optical to TeV photons, from thermal neutrons to heavy ions. Simulations can include any desired level of detail in three-dimensional geometries using the right level of detail in the reaction physics. The technology areas to which we have applied these codes include medical applications, defense, safety and security programs, nuclear safeguards and industrial and research system design and control. The main reason such applications are interesting is that by using these tools substantial savings of time and effort (i.e. money) can be realized. In addition it is possible to separate out and investigate computationally effects which can not be isolated and studied in experiments. In model calculations, just as in real life, one must take care in order to get the correct answer to the right question. Advancing computing technology allows extensions of Monte Carlo applications in two directions. First, as computers become more powerful more problems can be accurately modeled. Second, as computing power becomes cheaper Monte Carlo methods become accessible more widely. An overview of the set of Monte Carlo radiation transport tools in use a LLNL will be presented along with a few examples of applications and future directions
On the use of the Serpent Monte Carlo code for few-group cross section generation
International Nuclear Information System (INIS)
Fridman, E.; Leppaenen, J.
2011-01-01
Research highlights: → B1 methodology was used for generation of leakage-corrected few-group cross sections in the Serpent Monte-Carlo code. → Few-group constants generated by Serpent were compared with those calculated by Helios deterministic lattice transport code. → 3D analysis of a PWR core was performed by a nodal diffusion code DYN3D employing two-group cross section sets generated by Serpent and Helios. → An excellent agreement in the results of 3D core calculations obtained with Helios and Serpent generated cross-section libraries was observed. - Abstract: Serpent is a recently developed 3D continuous-energy Monte Carlo (MC) reactor physics burnup calculation code. Serpent is specifically designed for lattice physics applications including generation of homogenized few-group constants for full-core core simulators. Currently in Serpent, the few-group constants are obtained from the infinite-lattice calculations with zero neutron current at the outer boundary. In this study, in order to account for the non-physical infinite-lattice approximation, B1 methodology, routinely used by deterministic lattice transport codes, was considered for generation of leakage-corrected few-group cross sections in the Serpent code. A preliminary assessment of the applicability of the B1 methodology for generation of few-group constants in the Serpent code was carried out according to the following steps. Initially, the two-group constants generated by Serpent were compared with those calculated by Helios deterministic lattice transport code. Then, a 3D analysis of a Pressurized Water Reactor (PWR) core was performed by the nodal diffusion code DYN3D employing two-group cross section sets generated by Serpent and Helios. At this stage thermal-hydraulic (T-H) feedback was neglected. The DYN3D results were compared with those obtained from the 3D full core Serpent MC calculations. Finally, the full core DYN3D calculations were repeated taking into account T-H feedback and
Recent Developments of JAEA's Monte Carlo Code MVP for Reactor Physics Applications
Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa
2014-06-01
This paper describes the recent development status of a Monte Carlo code MVP developed at Japan Atomic Energy Agency. The basic features and capabilities of MVP are overviewed. In addition, new capabilities useful for reactor analysis are also described.
Data decomposition of Monte Carlo particle transport simulations via tally servers
International Nuclear Information System (INIS)
Romano, Paul K.; Siegel, Andrew R.; Forget, Benoit; Smith, Kord
2013-01-01
An algorithm for decomposing large tally data in Monte Carlo particle transport simulations is developed, analyzed, and implemented in a continuous-energy Monte Carlo code, OpenMC. The algorithm is based on a non-overlapping decomposition of compute nodes into tracking processors and tally servers. The former are used to simulate the movement of particles through the domain while the latter continuously receive and update tally data. A performance model for this approach is developed, suggesting that, for a range of parameters relevant to LWR analysis, the tally server algorithm should perform with minimal overhead on contemporary supercomputers. An implementation of the algorithm in OpenMC is then tested on the Intrepid and Titan supercomputers, supporting the key predictions of the model over a wide range of parameters. We thus conclude that the tally server algorithm is a successful approach to circumventing classical on-node memory constraints en route to unprecedentedly detailed Monte Carlo reactor simulations
International Nuclear Information System (INIS)
Hoogenboom, J.E.
1981-01-01
An adjoint Monte Carlo technique is described for the solution of neutron transport problems. The optimum biasing function for a zero-variance collision estimator is derived. The optimum treatment of an analog of a non-velocity thermal group has also been derived. The method is extended to multiplying systems, especially for eigenfunction problems to enable the estimate of averages over the unknown fundamental neutron flux distribution. A versatile computer code, FOCUS, has been written, based on the described theory. Numerical examples are given for a shielding problem and a critical assembly, illustrating the performance of the FOCUS code. 19 refs
A midway forward-adjoint coupling method for neutron and photon Monte Carlo transport
International Nuclear Information System (INIS)
Serov, I.V.; John, T.M.; Hoogenboom, J.E.
1999-01-01
The midway Monte Carlo method for calculating detector responses combines a forward and an adjoint Monte Carlo calculation. In both calculations, particle scores are registered at a surface to be chosen by the user somewhere between the source and detector domains. The theory of the midway response determination is developed within the framework of transport theory for external sources and for criticality theory. The theory is also developed for photons, which are generated at inelastic scattering or capture of neutrons. In either the forward or the adjoint calculation a so-called black absorber technique can be applied; i.e., particles need not be followed after passing the midway surface. The midway Monte Carlo method is implemented in the general-purpose MCNP Monte Carlo code. The midway Monte Carlo method is demonstrated to be very efficient in problems with deep penetration, small source and detector domains, and complicated streaming paths. All the problems considered pose difficult variance reduction challenges. Calculations were performed using existing variance reduction methods of normal MCNP runs and using the midway method. The performed comparative analyses show that the midway method appears to be much more efficient than the standard techniques in an overwhelming majority of cases and can be recommended for use in many difficult variance reduction problems of neutral particle transport
Modelling plastic scintillator response to gamma rays using light transport incorporated FLUKA code
Energy Technology Data Exchange (ETDEWEB)
Ranjbar Kohan, M. [Physics Department, Tafresh University, Tafresh (Iran, Islamic Republic of); Etaati, G.R. [Department of Nuclear Engineering and Physics, Amir Kabir University of Technology, Tehran (Iran, Islamic Republic of); Ghal-Eh, N., E-mail: ghal-eh@du.ac.ir [School of Physics, Damghan University, Damghan (Iran, Islamic Republic of); Safari, M.J. [Department of Energy Engineering, Sharif University of Technology, Tehran (Iran, Islamic Republic of); Afarideh, H. [Department of Nuclear Engineering and Physics, Amir Kabir University of Technology, Tehran (Iran, Islamic Republic of); Asadi, E. [Department of Physics, Payam-e-Noor University, Tehran (Iran, Islamic Republic of)
2012-05-15
The response function of NE102 plastic scintillator to gamma rays has been simulated using a joint FLUKA+PHOTRACK Monte Carlo code. The multi-purpose particle transport code, FLUKA, has been responsible for gamma transport whilst the light transport code, PHOTRACK, has simulated the transport of scintillation photons through scintillator and lightguide. The simulation results of plastic scintillator with/without light guides of different surface coverings have been successfully verified with experiments. - Highlights: Black-Right-Pointing-Pointer A multi-purpose code (FLUKA) and a light transport code (PHOTRACK) have been linked. Black-Right-Pointing-Pointer The hybrid code has been used to generate the response function of an NE102 scintillator. Black-Right-Pointing-Pointer The simulated response functions exhibit a good agreement with experimental data.
Monte Carlo simulation on nuclear energy study. Annual report of Nuclear Code Evaluation Committee
International Nuclear Information System (INIS)
Sakurai, Kiyoshi; Yamamoto, Toshihiro
1999-03-01
In this report, research results discussed in 1998 fiscal year at Nuclear Code Evaluation Special Committee of Nuclear Code Committee were summarised. Present status of Monte Carlo calculation in high energy region investigated / discussed at Monte Carlo simulation working-group and automatic compilation system for MCNP cross sections developed at MCNP high temperature library compilation working-group were described. The 6 papers are indexed individually. (J.P.N.)
Recent developments of JAEA’s Monte Carlo code MVP for reactor physics applications
International Nuclear Information System (INIS)
Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa
2015-01-01
Highlights: • This paper describes the recent development status of the Monte Carlo code MVP. • The basic features and capabilities of MVP are briefly described. • New capabilities useful for reactor analysis are also described. - Abstract: This paper describes the recent development status of a Monte Carlo code MVP developed at Japan Atomic Energy Agency. The basic features and capabilities of MVP are overviewed. In addition, new capabilities useful for reactor analysis are also described
International Nuclear Information System (INIS)
Wu, Xu; Kozlowski, Tomasz
2015-01-01
Highlights: • Coupling of Monte Carlo code Serpent and thermal–hydraulics code RELAP5. • A convergence criterion is developed based on the statistical uncertainty of power. • Correlation between MC statistical uncertainty and coupled error is quantified. • Both UO 2 and MOX single assembly models are used in the coupled simulation. • Validation of coupling results with a multi-group transport code DeCART. - Abstract: Coupled multi-physics approach plays an important role in improving computational accuracy. Compared with deterministic neutronics codes, Monte Carlo codes have the advantage of a higher resolution level. In the present paper, a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, Serpent, is coupled with a thermal–hydraulics safety analysis code, RELAP5. The coupled Serpent/RELAP5 code capability is demonstrated by the improved axial power distribution of UO 2 and MOX single assembly models, based on the OECD-NEA/NRC PWR MOX/UO 2 Core Transient Benchmark. Comparisons of calculation results using the coupled code with those from the deterministic methods, specifically heterogeneous multi-group transport code DeCART, show that the coupling produces more precise results. A new convergence criterion for the coupled simulation is developed based on the statistical uncertainty in power distribution in the Monte Carlo code, rather than ad-hoc criteria used in previous research. The new convergence criterion is shown to be more rigorous, equally convenient to use but requiring a few more coupling steps to converge. Finally, the influence of Monte Carlo statistical uncertainty on the coupled error of power and thermal–hydraulics parameters is quantified. The results are presented such that they can be used to find the statistical uncertainty to use in Monte Carlo in order to achieve a desired precision in coupled simulation
Analysis of error in Monte Carlo transport calculations
International Nuclear Information System (INIS)
Booth, T.E.
1979-01-01
The Monte Carlo method for neutron transport calculations suffers, in part, because of the inherent statistical errors associated with the method. Without an estimate of these errors in advance of the calculation, it is difficult to decide what estimator and biasing scheme to use. Recently, integral equations have been derived that, when solved, predicted errors in Monte Carlo calculations in nonmultiplying media. The present work allows error prediction in nonanalog Monte Carlo calculations of multiplying systems, even when supercritical. Nonanalog techniques such as biased kernels, particle splitting, and Russian Roulette are incorporated. Equations derived here allow prediction of how much a specific variance reduction technique reduces the number of histories required, to be weighed against the change in time required for calculation of each history. 1 figure, 1 table
Running the EGS4 Monte Carlo code with Fortran 90 on a pentium computer
International Nuclear Information System (INIS)
Caon, M.; Bibbo, G.; Pattison, J.
1996-01-01
The possibility to run the EGS4 Monte Carlo code radiation transport system for medical radiation modelling on a microcomputer is discussed. This has been done using a Fortran 77 compiler with a 32-bit memory addressing system running under a memory extender operating system. In addition a virtual memory manager such as QEMM386 was required. It has successfully run on a SUN Sparcstation2. In 1995 faster Pentium-based microcomputers became available as did the Windows 95 operating system which can handle 32-bit programs, multitasking and provides its own virtual memory management. The paper describe how with simple modification to the batch files it was possible to run EGS4 on a Pentium under Fortran 90 and Windows 95. This combination of software and hardware is cheaper and faster than running it on a SUN Sparcstation2. 8 refs., 1 tab
Running the EGS4 Monte Carlo code with Fortran 90 on a pentium computer
Energy Technology Data Exchange (ETDEWEB)
Caon, M. [Flinders Univ. of South Australia, Bedford Park, SA (Australia)]|[Univercity of South Australia, SA (Australia); Bibbo, G. [Womens and Childrens hospital, SA (Australia); Pattison, J. [Univercity of South Australia, SA (Australia)
1996-09-01
The possibility to run the EGS4 Monte Carlo code radiation transport system for medical radiation modelling on a microcomputer is discussed. This has been done using a Fortran 77 compiler with a 32-bit memory addressing system running under a memory extender operating system. In addition a virtual memory manager such as QEMM386 was required. It has successfully run on a SUN Sparcstation2. In 1995 faster Pentium-based microcomputers became available as did the Windows 95 operating system which can handle 32-bit programs, multitasking and provides its own virtual memory management. The paper describe how with simple modification to the batch files it was possible to run EGS4 on a Pentium under Fortran 90 and Windows 95. This combination of software and hardware is cheaper and faster than running it on a SUN Sparcstation2. 8 refs., 1 tab.
OpenMC: a state-of-the-Art Monte Carlo code for research and development
International Nuclear Information System (INIS)
Romano, P.K.; Horelik, N.E.; Herman, B.R.; Forget, B.; Smith, K.; Nelson, A.G.
2013-01-01
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. (authors)
Monte Carlo simulations of the particle transport in semiconductor detectors of fast neutrons
International Nuclear Information System (INIS)
Sedlačková, Katarína; Zaťko, Bohumír; Šagátová, Andrea; Nečas, Vladimír
2013-01-01
Several Monte Carlo all-particle transport codes are under active development around the world. In this paper we focused on the capabilities of the MCNPX code (Monte Carlo N-Particle eXtended) to follow the particle transport in semiconductor detector of fast neutrons. Semiconductor detector based on semi-insulating GaAs was the object of our investigation. As converter material capable to produce charged particles from the (n, p) interaction, a high-density polyethylene (HDPE) was employed. As the source of fast neutrons, the 239 Pu–Be neutron source was used in the model. The simulations were performed using the MCNPX code which makes possible to track not only neutrons but also recoiled protons at all interesting energies. Hence, the MCNPX code enables seamless particle transport and no other computer program is needed to process the particle transport. The determination of the optimal thickness of the conversion layer and the minimum thickness of the active region of semiconductor detector as well as the energy spectra simulation were the principal goals of the computer modeling. Theoretical detector responses showed that the best detection efficiency can be achieved for 500 μm thick HDPE converter layer. The minimum detector active region thickness has been estimated to be about 400 μm. -- Highlights: ► Application of the MCNPX code for fast neutron detector design is demonstrated. ► Simulations of the particle transport through conversion film of HDPE are presented. ► Simulations of the particle transport through detector active region are presented. ► The optimal thickness of the HDPE conversion film has been calculated. ► Detection efficiency of 0.135% was reached for 500 μm thick HDPE conversion film
International Nuclear Information System (INIS)
Bellezzo, Murillo
2014-01-01
As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo Method (MCM) has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this thesis, the CUBMC code is presented, a GPU-based MC photon transport algorithm for dose calculation under the Compute Unified Device Architecture (CUDA) platform. The simulation of physical events is based on the algorithm used in PENELOPE, and the cross section table used is the one generated by the MATERIAL routine, also present in PENELOPE code. Photons are transported in voxel-based geometries with different compositions. There are two distinct approaches used for transport simulation. The rst of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon ignores the existence of borders and travels in homogeneous fictitious media. The CUBMC code aims to be an alternative of Monte Carlo simulator code that, by using the capability of parallel processing of graphics processing units (GPU), provide high performance simulations in low cost compact machines, and thus can be applied in clinical cases and incorporated in treatment planning systems for radiotherapy. (author)
Development of M3C code for Monte Carlo reactor physics criticality calculations
International Nuclear Information System (INIS)
Kumar, Anek; Kannan, Umasankari; Krishanani, P.D.
2015-06-01
The development of Monte Carlo code (M3C) for reactor design entails use of continuous energy nuclear data and Monte Carlo simulations for each of the neutron interaction processes. BARC has started a concentrated effort for developing a new general geometry continuous energy Monte Carlo code for reactor physics calculation indigenously. The code development required a comprehensive understanding of the basic continuous energy cross section sets. The important features of this code are treatment of heterogeneous lattices by general geometry, use of point cross sections along with unionized energy grid approach, thermal scattering model for low energy treatment, capability of handling the microscopic fuel particles dispersed randomly. The capability of handling the randomly dispersed microscopic fuel particles which is very useful for the modeling of High-Temperature Gas-Cooled reactor fuels which are composed of thousands of microscopic fuel particle (TRISO fuel particle), randomly dispersed in a graphite matrix. The Monte Carlo code for criticality calculation is a pioneering effort and has been used to study several types of lattices including cluster geometries. The code has been verified for its accuracy against more than 60 sample problems covering a wide range from simple (like spherical) to complex geometry (like PHWR lattice). Benchmark results show that the code performs quite well for the criticality calculation of the system. In this report, the current status of the code, features of the code, some of the benchmark results for the testing of the code and input preparation etc. are discussed. (author)
Generic programming for deterministic neutron transport codes
International Nuclear Information System (INIS)
Plagne, L.; Poncot, A.
2005-01-01
This paper discusses the implementation of neutron transport codes via generic programming techniques. Two different Boltzmann equation approximations have been implemented, namely the Sn and SPn methods. This implementation experiment shows that generic programming allows us to improve maintainability and readability of source codes with no performance penalties compared to classical approaches. In the present implementation, matrices and vectors as well as linear algebra algorithms are treated separately from the rest of source code and gathered in a tool library called 'Generic Linear Algebra Solver System' (GLASS). Such a code architecture, based on a linear algebra library, allows us to separate the three different scientific fields involved in transport codes design: numerical analysis, reactor physics and computer science. Our library handles matrices with optional storage policies and thus applies both to Sn code, where the matrix elements are computed on the fly, and to SPn code where stored matrices are used. Thus, using GLASS allows us to share a large fraction of source code between Sn and SPn implementations. Moreover, the GLASS high level of abstraction allows the writing of numerical algorithms in a form which is very close to their textbook descriptions. Hence the GLASS algorithms collection, disconnected from computer science considerations (e.g. storage policy), is very easy to read, to maintain and to extend. (authors)
International Nuclear Information System (INIS)
Malatara, G.; Kappas, K.; Sphiris, N.
1994-01-01
In this work, a Monte Carlo EGS4 code was used to simulate radiation transport through linear accelerators to produce and score energy spectra and angular distributions of 6, 12, 15 and 25 MeV bremsstrahlung photons exiting from different accelerator treatment heads. The energy spectra was used as input for a convolution method program to calculate the tissue-maximum ratio in water. 100.000 histories are recorded in the scoring plane for each simulation. The validity of the Monte Carlo simulation and the precision to the calculated spectra have been verified experimentally and were in good agreement. We believe that the accurate simulation of the different components of the linear accelerator head is very important for the precision of the results. The results of the Monte Carlo and the Convolution Method can be compared with experimental data for verification and they are powerful and practical tools to generate accurate spectra and dosimetric data. (authors)
International Nuclear Information System (INIS)
Walsh, Jonathan A.; Palmer, Todd S.; Urbatsch, Todd J.
2015-01-01
Highlights: • Generation of discrete differential scattering angle and energy loss cross sections. • Gauss–Radau quadrature utilizing numerically computed cross section moments. • Development of a charged particle transport capability in the Milagro IMC code. • Integration of cross section generation and charged particle transport capabilities. - Abstract: We investigate a method for numerically generating discrete scattering cross sections for use in charged particle transport simulations. We describe the cross section generation procedure and compare it to existing methods used to obtain discrete cross sections. The numerical approach presented here is generalized to allow greater flexibility in choosing a cross section model from which to derive discrete values. Cross section data computed with this method compare favorably with discrete data generated with an existing method. Additionally, a charged particle transport capability is demonstrated in the time-dependent Implicit Monte Carlo radiative transfer code, Milagro. We verify the implementation of charged particle transport in Milagro with analytic test problems and we compare calculated electron depth–dose profiles with another particle transport code that has a validated electron transport capability. Finally, we investigate the integration of the new discrete cross section generation method with the charged particle transport capability in Milagro.
Benchmarking time-dependent neutron problems with Monte Carlo codes
International Nuclear Information System (INIS)
Couet, B.; Loomis, W.A.
1990-01-01
Many nuclear logging tools measure the time dependence of a neutron flux in a geological formation to infer important properties of the formation. The complex geometry of the tool and the borehole within the formation does not permit an exact deterministic modelling of the neutron flux behaviour. While this exact simulation is possible with Monte Carlo methods the computation time does not facilitate quick turnaround of results useful for design and diagnostic purposes. Nonetheless a simple model based on the diffusion-decay equation for the flux of neutrons of a single energy group can be useful in this situation. A combination approach where a Monte Carlo calculation benchmarks a deterministic model in terms of the diffusion constants of the neutrons propagating in the media and their flux depletion rates thus offers the possibility of quick calculation with assurance as to accuracy. We exemplify this approach with the Monte Carlo benchmarking of a logging tool problem, showing standoff and bedding response. (author)
Vectorization and parallelization of Monte-Carlo programs for calculation of radiation transport
International Nuclear Information System (INIS)
Seidel, R.
1995-01-01
The versatile MCNP-3B Monte-Carlo code written in FORTRAN77, for simulation of the radiation transport of neutral particles, has been subjected to vectorization and parallelization of essential parts, without touching its versatility. Vectorization is not dependent on a specific computer. Several sample tasks have been selected in order to test the vectorized MCNP-3B code in comparison to the scalar MNCP-3B code. The samples are a representative example of the 3-D calculations to be performed for simulation of radiation transport in neutron and reactor physics. (1) 4πneutron detector. (2) High-energy calorimeter. (3) PROTEUS benchmark (conversion rates and neutron multiplication factors for the HCLWR (High Conversion Light Water Reactor)). (orig./HP) [de
Coupled geochemical and solute transport code development
International Nuclear Information System (INIS)
Morrey, J.R.; Hostetler, C.J.
1985-01-01
A number of coupled geochemical hydrologic codes have been reported in the literature. Some of these codes have directly coupled the source-sink term to the solute transport equation. The current consensus seems to be that directly coupling hydrologic transport and chemical models through a series of interdependent differential equations is not feasible for multicomponent problems with complex geochemical processes (e.g., precipitation/dissolution reactions). A two-step process appears to be the required method of coupling codes for problems where a large suite of chemical reactions must be monitored. Two-step structure requires that the source-sink term in the transport equation is supplied by a geochemical code rather than by an analytical expression. We have developed a one-dimensional two-step coupled model designed to calculate relatively complex geochemical equilibria (CTM1D). Our geochemical module implements a Newton-Raphson algorithm to solve heterogeneous geochemical equilibria, involving up to 40 chemical components and 400 aqueous species. The geochemical module was designed to be efficient and compact. A revised version of the MINTEQ Code is used as a parent geochemical code
Implementation of 3D models in the Monte Carlo code MCNP
International Nuclear Information System (INIS)
Lopes, Vivaldo; Millian, Felix M.; Guevara, Maria Victoria M.; Garcia, Fermin; Sena, Isaac; Menezes, Hugo
2009-01-01
On the area of numerical dosimetry Applied to medical physics, the scientific community focuses on the elaboration of new hybrids models based on 3D models. But different steps of the process of simulation with 3D models needed improvement and optimization in order to expedite the calculations and accuracy using this methodology. This project was developed with the aim of optimize the process of introduction of 3D models within the simulation code of radiation transport by Monte Carlo (MCNP). The fast implementation of these models on the simulation code allows the estimation of the dose deposited on the patient organs on a more personalized way, increasing the accuracy with this on the estimates and reducing the risks to health, caused by ionizing radiations. The introduction o these models within the MCNP was made through a input file, that was constructed through a sequence of images, bi-dimensional in the 3D model, generated using the program '3DSMAX', imported by the program 'TOMO M C' and thus, introduced as INPUT FILE of the MCNP code. (author)
International Nuclear Information System (INIS)
Quade, U.
1994-01-01
Neutron- und Gamma dose rate calculations were performed for the storage containers filled with plutonium nitrate of the MOX fabrication facility of Siemens. For the particle transport calculations the Monte Carlo Code MCNP 4.2 was used. The calculated results were compared with experimental dose rate measurements. It can be stated that the choice of the code system was appropriate since all aspects of the many facettes of the problem were well reproduced in the calculations. The position dependency as well as the influence of the shieldings, the reflections and the mutual influences of the sources were well described by the calculations for the gamma and for the neutron dose rates. However, good agreement with the experimental results on the gamma dose rates could only be reached when the lead shielding of the detector was integrated into the geometry modelling of the calculations. For some few cases of thick shieldings and soft gamma ray sources the statistics of the calculational results were not sufficient. In such cases more elaborate variance reduction methods must be applied in future calculations. Thus the MCNP code in connection with NGSRC has been proven as an effective tool for the solution of this type of problems. (orig./HP) [de
Recent R and D around the Monte-Carlo code Tripoli-4 for criticality calculation
International Nuclear Information System (INIS)
Hugot, F.X.; Lee, Y.K.; Malvagi, F.
2008-01-01
TRIPOLI-4 [1] is the fourth generation of the TRIPOLI family of Monte Carlo codes developed from the 60's by CEA. It simulates the 3D transport of neutrons, photons, electrons and positrons as well as coupled neutron-photon propagation and electron-photons cascade showers. The code addresses radiation protection and shielding problems, as well as criticality and reactor physics problems through both critical and subcritical neutronics calculations. It uses full pointwise as well as multigroup cross-sections. The code has been validated through several hundred benchmarks as well as measurement campaigns. It is used as a reference tool by CEA as well as its industrial and institutional partners, and in the NURESIM [2] European project. Section 2 reviews its main features, with emphasis on the latest developments. Section 3 presents some recent R and D for criticality calculations. Fission matrix, Eigen-values and eigenvectors computations will be exposed. Corrections on the standard deviation estimator in the case of correlations between generation steps will be detailed. Section 4 presents some preliminary results obtained by the new mesh tally feature. The last section presents the interest of using XML format output files. (authors)
A perturbation-based susbtep method for coupled depletion Monte-Carlo codes
International Nuclear Information System (INIS)
Kotlyar, Dan; Aufiero, Manuele; Shwageraus, Eugene; Fratoni, Massimiliano
2017-01-01
Highlights: • The GPT method allows to calculate the sensitivity coefficients to any perturbation. • Full Jacobian of sensitivities, cross sections (XS) to concentrations, may be obtained. • The time dependent XS is obtained by combining the GPT and substep methods. • The proposed GPT substep method considerably reduces the time discretization error. • No additional MC transport solutions are required within the time step. - Abstract: Coupled Monte Carlo (MC) methods are becoming widely used in reactor physics analysis and design. Many research groups therefore, developed their own coupled MC depletion codes. Typically, in such coupled code systems, neutron fluxes and cross sections are provided to the depletion module by solving a static neutron transport problem. These fluxes and cross sections are representative only of a specific time-point. In reality however, both quantities would change through the depletion time interval. Recently, Generalized Perturbation Theory (GPT) equivalent method that relies on collision history approach was implemented in Serpent MC code. This method was used here to calculate the sensitivity of each nuclide and reaction cross section due to the change in concentration of every isotope in the system. The coupling method proposed in this study also uses the substep approach, which incorporates these sensitivity coefficients to account for temporal changes in cross sections. As a result, a notable improvement in time dependent cross section behavior was obtained. The method was implemented in a wrapper script that couples Serpent with an external depletion solver. The performance of this method was compared with other existing methods. The results indicate that the proposed method requires substantially less MC transport solutions to achieve the same accuracy.
Intracoin - International Nuclide Transport Code Intercomparison Study
International Nuclear Information System (INIS)
1984-09-01
The purpose of the project is to obtain improved knowledge of the influence of various strategies for radionuclide transport modelling for the safety assessment of final repositories for nuclear waste. This is a report of the first phase of the project which was devoted to a comparison of the numerical accuracy of the computer codes used in the study. The codes can be divided into five groups, namely advection-dispersion models, models including matrix diffusion and chemical effects and finally combined models. The results are presented as comparisons of calculations since the objective of level 1 was code verification. (G.B.)
The Monte Carlo SRNA-VOX code for 3D proton dose distribution in voxelized geometry using CT data
International Nuclear Information System (INIS)
Ilic, Radovan D; Spasic-Jokic, Vesna; Belicev, Petar; Dragovic, Milos
2005-01-01
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
Evaluation and comparison of SN and Monte-Carlo charged particle transport calculations
International Nuclear Information System (INIS)
Hadad, K.
2000-01-01
A study was done to evaluate a 3-D S N charged particle transport code called SMARTEPANTS 1 and another 3-D Monte Carlo code called Integrated Tiger Series, ITS 2 . The evaluation study of SMARTEPANTS code was based on angular discretization and reflected boundary sensitivity whilst the evaluation of ITS was based on CPU time and variance reduction. The comparison of the two code was based on energy and charge deposition calculation in block of Gallium Arsenide with embedded gold cylinders. The result of evaluation tests shows that an S 8 calculation maintains both accuracy and speed and calculations with reflected boundaries geometry produces full symmetrical results. As expected for ITS evaluation, the CPU time and variance reduction are opposite to a point beyond which the history augmentation while increasing the CPU time do not result in variance reduction. The comparison test problem showed excellent agreement in total energy deposition calculations
Implementation of an approximate zero-variance scheme in the TRIPOLI Monte Carlo code
Energy Technology Data Exchange (ETDEWEB)
Christoforou, S.; Hoogenboom, J. E. [Delft Univ. of Technology, Mekelweg 15, 2629 JB Delft (Netherlands); Dumonteil, E.; Petit, O.; Diop, C. [Commissariat a l' Energie Atomique CEA, Gif-sur-Yvette (France)
2006-07-01
In an accompanying paper it is shown that theoretically a zero-variance Monte Carlo scheme can be devised for criticality calculations if the space, energy and direction dependent adjoint function is exactly known. This requires biasing of the transition and collision kernels with the appropriate adjoint function. In this paper it is discussed how an existing general purpose Monte Carlo code like TRIPOLI can be modified to approach the zero-variance scheme. This requires modifications for reading in the adjoint function obtained from a separate deterministic calculation for a number of space intervals, energy groups and discrete directions. Furthermore, a function has to be added to supply the direction dependent and the averaged adjoint function at a specific position in the system by interpolation. The initial particle weights of a certain batch must be set inversely proportional to the averaged adjoint function and proper normalization of the initial weights must be secured. The sampling of the biased transition kernel requires cumulative integrals of the biased kernel along the flight path until a certain value, depending on a selected random number is reached to determine a new collision site. The weight of the particle must be adapted accordingly. The sampling of the biased collision kernel (in a multigroup treatment) is much more like the normal sampling procedure. A numerical example is given for a 3-group calculation with a simplified transport model (two-direction model), demonstrating that the zero-variance scheme can be approximated quite well for this simplified case. (authors)
Radiosteoplasty study in animal bone and radiodosimetric evaluation using Monte Carlo code
Energy Technology Data Exchange (ETDEWEB)
Silveira, Marcia Flavia; Campos, Tarcisio Passos Ribeiro [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: marciaflaviafisio@gmail.com; campos@nuclear.ufmg.br
2007-07-01
The radiosteoplasty is a procedure that consists of the injection of a radioactive biomaterial incorporated to the bone cement into the osseous structure affected by cancer. This technique has been developed with the major objective to control the tumor or the regional bone metastasis (in situ) besides pain reduction and structural resistance increasing. In the present study the radiosteoplasty is applied to the bovine and swine bones in vitro using non-radioactive cement. The objective is to know the spatial distribution of the cold compound (non radioactive) in pig and ox bones after implant. A 2 mm needle was introduced into the cortical bone previously perforated. The distribution of this biomaterial was observed trough radiological images obtained just after the compound application. Recent dosimetric studies using Monte Carlo N-Particle method (MCNP-5) concluded that the spatial dose distribution is suitable for the protocol namely radiosteoplasty applied to treat bone tumors on superior and inferior members. The Monte Carlo method simulates the present process and it is particularly interesting tool to solve the complex photon and electron particle transport problems that can not be modeled by codes based on deterministic methods. These related radiodosimetric studies are presented and discussed. (author)
Regional Atmospheric Transport Code for Hanford Emission Tracking (RATCHET)
International Nuclear Information System (INIS)
Ramsdell, J.V. Jr.; Simonen, C.A.; Burk, K.W.
1994-02-01
The purpose of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate radiation doses that individuals may have received from operations at the Hanford Site since 1944. This report deals specifically with the atmospheric transport model, Regional Atmospheric Transport Code for Hanford Emission Tracking (RATCHET). RATCHET is a major rework of the MESOILT2 model used in the first phase of the HEDR Project; only the bookkeeping framework escaped major changes. Changes to the code include (1) significant changes in the representation of atmospheric processes and (2) incorporation of Monte Carlo methods for representing uncertainty in input data, model parameters, and coefficients. To a large extent, the revisions to the model are based on recommendations of a peer working group that met in March 1991. Technical bases for other portions of the atmospheric transport model are addressed in two other documents. This report has three major sections: a description of the model, a user's guide, and a programmer's guide. These sections discuss RATCHET from three different perspectives. The first provides a technical description of the code with emphasis on details such as the representation of the model domain, the data required by the model, and the equations used to make the model calculations. The technical description is followed by a user's guide to the model with emphasis on running the code. The user's guide contains information about the model input and output. The third section is a programmer's guide to the code. It discusses the hardware and software required to run the code. The programmer's guide also discusses program structure and each of the program elements
Modelling of the RA-1 reactor using a Monte Carlo code
International Nuclear Information System (INIS)
Quinteiro, Guillermo F.; Calabrese, Carlos R.
2000-01-01
It was carried out for the first time, a model of the Argentine RA-1 reactor using the MCNP Monte Carlo code. This model was validated using data for experimental neutron and gamma measurements at different energy ranges and locations. In addition, the resulting fluxes were compared with the data obtained using a 3D diffusion code. (author)
Discrete-ordinates electron transport calculations using standard neutron transport codes
International Nuclear Information System (INIS)
Morel, J.E.
1979-01-01
The primary purpose of this work was to develop a method for using standard neutron transport codes to perform electron transport calculations. The method is to develop approximate electron cross sections which are sufficiently well-behaved to be treated with standard S/sub n/ methods, but which nonetheless yield flux solutions which are very similar to the exact solutions. The main advantage of this approach is that, once the approximate cross sections are constructed, their multigroup Legendre expansion coefficients can be calculated and input to any standard S/sub n/ code. Discrete-ordinates calculations were performed to determine the accuracy of the flux solutions for problems corresponding to 1.0-MeV electrons incident upon slabs of aluminum and gold. All S/sub n/ calculations were compared with similar calculations performed with an electron Monte Carlo code, considered to be exact. In all cases, the discrete-ordinates solutions for integral flux quantities (i.e., scalar flux, energy deposition profiles, etc.) are generally in agreement with the Monte Carlo solutions to within approximately 5% or less. The central conclusion is that integral electron flux quantities can be efficiently and accurately calculated using standard S/sub n/ codes in conjunction with approximate cross sections. Furthermore, if group structures and approximate cross section construction are optimized, accurate differential flux energy spectra may also be obtainable without having to use an inordinately large number of energy groups. 1 figure
Monte Carlo analysis of radiative transport in oceanographic lidar measurements
Energy Technology Data Exchange (ETDEWEB)
Cupini, E.; Ferro, G. [ENEA, Divisione Fisica Applicata, Centro Ricerche Ezio Clementel, Bologna (Italy); Ferrari, N. [Bologna Univ., Bologna (Italy). Dipt. Ingegneria Energetica, Nucleare e del Controllo Ambientale
2001-07-01
The analysis of oceanographic lidar systems measurements is often carried out with semi-empirical methods, since there is only a rough understanding of the effects of many environmental variables. The development of techniques for interpreting the accuracy of lidar measurements is needed to evaluate the effects of various environmental situations, as well as of different experimental geometric configurations and boundary conditions. A Monte Carlo simulation model represents a tool that is particularly well suited for answering these important questions. The PREMAR-2F Monte Carlo code has been developed taking into account the main molecular and non-molecular components of the marine environment. The laser radiation interaction processes of diffusion, re-emission, refraction and absorption are treated. In particular are considered: the Rayleigh elastic scattering, produced by atoms and molecules with small dimensions with respect to the laser emission wavelength (i.e. water molecules), the Mie elastic scattering, arising from atoms or molecules with dimensions comparable to the laser wavelength (hydrosols), the Raman inelastic scattering, typical of water, the absorption of water, inorganic (sediments) and organic (phytoplankton and CDOM) hydrosols, the fluorescence re-emission of chlorophyll and yellow substances. PREMAR-2F is an extension of a code for the simulation of the radiative transport in atmospheric environments (PREMAR-2). The approach followed in PREMAR-2 was to combine conventional Monte Carlo techniques with analytical estimates of the probability of the receiver to have a contribution from photons coming back after an interaction in the field of view of the lidar fluorosensor collecting apparatus. This offers an effective mean for modelling a lidar system with realistic geometric constraints. The retrieved semianalytic Monte Carlo radiative transfer model has been developed in the frame of the Italian Research Program for Antarctica (PNRA) and it is
FLUKA: A Multi-Particle Transport Code
Energy Technology Data Exchange (ETDEWEB)
Ferrari, A.; Sala, P.R.; /CERN /INFN, Milan; Fasso, A.; /SLAC; Ranft, J.; /Siegen U.
2005-12-14
This report describes the 2005 version of the Fluka particle transport code. The first part introduces the basic notions, describes the modular structure of the system, and contains an installation and beginner's guide. The second part complements this initial information with details about the various components of Fluka and how to use them. It concludes with a detailed history and bibliography.
Mendenhall, Marcus H.; Weller, Robert A.
2011-01-01
In Monte Carlo particle transport codes, it is often important to adjust reaction cross sections to reduce the variance of calculations of relatively rare events, in a technique known as non-analogous Monte Carlo. We present the theory and sample code for a Geant4 process which allows the cross section of a G4VDiscreteProcess to be scaled, while adjusting track weights so as to mitigate the effects of altered primary beam depletion induced by the cross section change. This makes it possible t...
Colloid transport code-nuclear user's manual
International Nuclear Information System (INIS)
Jain, R.
1992-01-01
This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential systems
RITA, a promising Monte Carlo code for recoil implantation
International Nuclear Information System (INIS)
Desalvo, A.; Rosa, R.
1982-01-01
A computer code previously set up to simulate ion penetration in amorphous solids has been extended to handle with recoil phenomena. Preliminary results are compared with existing experimental data. (author)
Jetto a free boundary plasma transport code
International Nuclear Information System (INIS)
Cenacchi, G.; Taroni, A.
1988-01-01
JETTO is a one-and-a-half-dimensional transport code calculating the evolution of plasma parameters in a time dependent axisymmetric MHD equilibrium configuration. A splitting technique gives a consistent solution of coupled equilibrium and transport equations. The plasma boundary is free and defined either by its contact with a limiter (wall) or by a separatrix or by the toroidal magnetic flux. The Grad's approach to the equilibrium problem with adiabatic (or similar) constraints is adopted. This method consists of iterating by alternately solving the Grad-Schluter-Shafranov equation (PDE) and the ODE obtained by averaging the PDE over the magnetic surfaces. The bidimensional equation of the poloidal flux is solved by a finite difference scheme, whereas a Runge-Kutta method is chosen for the averaged equilibrium equation. The 1D transport equations (averaged over the magnetic surfaces) for the electron and ion densities and energies and for the rotational transform are written in terms of a coordinate (ρ) related to the toroidal flux. Impurity transport is also considered, under the hypothesis of coronal equilibrium. The transport equations are solved by an implicit scheme in time and by a finite difference scheme in space. The centering of the source terms and transport coefficients is performed using a Predictor-Corrector scheme. The basic version of the code is described here in detail; input and output parameters are also listed
Monte carlo depletion analysis of SMART core by MCNAP code
International Nuclear Information System (INIS)
Jung, Jong Sung; Sim, Hyung Jin; Kim, Chang Hyo; Lee, Jung Chan; Ji, Sung Kyun
2001-01-01
Depletion an analysis of SMART, a small-sized advanced integral PWR under development by KAERI, is conducted using the Monte Carlo (MC) depletion analysis program, MCNAP. The results are compared with those of the CASMO-3/ MASTER nuclear analysis. The difference between MASTER and MCNAP on k eff prediction is observed about 600pcm at BOC, and becomes smaller as the core burnup increases. The maximum difference bet ween two predict ions on fuel assembly (FA) normalized power distribution is about 6.6% radially , and 14.5% axially but the differences are observed to lie within standard deviation of MC estimations
Monitoring and preventing numerical oscillations in 3D simulations with coupled Monte Carlo codes
International Nuclear Information System (INIS)
Kotlyar, D.; Shwageraus, E.
2014-01-01
Highlights: • Conventional coupling methods used in all MC codes can be numerically unstable. • Application of new stochastic implicit (SIMP) methods may be required. • The implicit methods require additional computational effort. • Monitoring diagnostic of the numerical stability was developed here. • The procedure allows to create an hybrid explicit–implicit coupling scheme. - Abstract: Previous studies have reported that different schemes for coupling Monte Carlo (MC) neutron transport with burnup and thermal hydraulic feedbacks may potentially be numerically unstable. This issue can be resolved by application of implicit methods, such as the stochastic implicit mid-point (SIMP) methods. In order to assure numerical stability, the new methods do require additional computational effort. The instability issue however, is problem-dependent and does not necessarily occur in all cases. Therefore, blind application of the unconditionally stable coupling schemes, and thus incurring extra computational costs, may not always be necessary. In this paper, we attempt to develop an intelligent diagnostic mechanism, which will monitor numerical stability of the calculations and, if necessary, switch from simple and fast coupling scheme to more computationally expensive but unconditionally stable one. To illustrate this diagnostic mechanism, we performed a coupled burnup and TH analysis of a single BWR fuel assembly. The results indicate that the developed algorithm can be easily implemented in any MC based code for monitoring of numerical instabilities. The proposed monitoring method has negligible impact on the calculation time even for realistic 3D multi-region full core calculations
First validation of the new continuous energy version of the MORET5 Monte Carlo code
International Nuclear Information System (INIS)
Miss, Joachim; Bernard, Franck; Forestier, Benoit; Haeck, Wim; Richet, Yann; Jacquet, Olivier
2008-01-01
The 5.A.1 version is the next release of the MORET Monte Carlo code dedicated to criticality and reactor calculations. This new version combines all the capabilities that are already available in the multigroup version with many new and enhanced features. The main capabilities of the previous version are the powerful association of a deterministic and Monte Carlo approach (like for instance APOLLO-MORET), the modular geometry, five source sampling techniques and two simulation strategies. The major advance in MORET5 is the ability to perform calculations either a multigroup or a continuous energy simulation. Thanks to these new developments, we now have better control over the whole process of criticality calculations, from reading the basic nuclear data to the Monte Carlo simulation itself. Moreover, this new capability enables us to better validate the deterministic-Monte Carlo multigroup calculations by performing continuous energy calculations with the same code, using the same geometry and tracking algorithms. The aim of this paper is to describe the main options available in this new release, and to present the first results. Comparisons of the MORET5 continuous-energy results with experimental measurements and against another continuous-energy Monte Carlo code are provided in terms of validation and time performance. Finally, an analysis of the interest of using a unified energy grid for continuous energy Monte Carlo calculations is presented. (authors)
Energy Technology Data Exchange (ETDEWEB)
Dixon, D.A., E-mail: ddixon@lanl.gov [Los Alamos National Laboratory, P.O. Box 1663, MS P365, Los Alamos, NM 87545 (United States); Prinja, A.K., E-mail: prinja@unm.edu [Department of Nuclear Engineering, MSC01 1120, 1 University of New Mexico, Albuquerque, NM 87131-0001 (United States); Franke, B.C., E-mail: bcfrank@sandia.gov [Sandia National Laboratories, Albuquerque, NM 87123 (United States)
2015-09-15
This paper presents the theoretical development and numerical demonstration of a moment-preserving Monte Carlo electron transport method. Foremost, a full implementation of the moment-preserving (MP) method within the Geant4 particle simulation toolkit is demonstrated. Beyond implementation details, it is shown that the MP method is a viable alternative to the condensed history (CH) method for inclusion in current and future generation transport codes through demonstration of the key features of the method including: systematically controllable accuracy, computational efficiency, mathematical robustness, and versatility. A wide variety of results common to electron transport are presented illustrating the key features of the MP method. In particular, it is possible to achieve accuracy that is statistically indistinguishable from analog Monte Carlo, while remaining up to three orders of magnitude more efficient than analog Monte Carlo simulations. Finally, it is shown that the MP method can be generalized to any applicable analog scattering DCS model by extending previous work on the MP method beyond analytical DCSs to the partial-wave (PW) elastic tabulated DCS data.
The Monte Carlo photoionization and moving-mesh radiation hydrodynamics code CMACIONIZE
Vandenbroucke, B.; Wood, K.
2018-04-01
We present the public Monte Carlo photoionization and moving-mesh radiation hydrodynamics code CMACIONIZE, which can be used to simulate the self-consistent evolution of HII regions surrounding young O and B stars, or other sources of ionizing radiation. The code combines a Monte Carlo photoionization algorithm that uses a complex mix of hydrogen, helium and several coolants in order to self-consistently solve for the ionization and temperature balance at any given type, with a standard first order hydrodynamics scheme. The code can be run as a post-processing tool to get the line emission from an existing simulation snapshot, but can also be used to run full radiation hydrodynamical simulations. Both the radiation transfer and the hydrodynamics are implemented in a general way that is independent of the grid structure that is used to discretize the system, allowing it to be run both as a standard fixed grid code, but also as a moving-mesh code.
Condensed history Monte Carlo methods for photon transport problems
International Nuclear Information System (INIS)
Bhan, Katherine; Spanier, Jerome
2007-01-01
We study methods for accelerating Monte Carlo simulations that retain most of the accuracy of conventional Monte Carlo algorithms. These methods - called Condensed History (CH) methods - have been very successfully used to model the transport of ionizing radiation in turbid systems. Our primary objective is to determine whether or not such methods might apply equally well to the transport of photons in biological tissue. In an attempt to unify the derivations, we invoke results obtained first by Lewis, Goudsmit and Saunderson and later improved by Larsen and Tolar. We outline how two of the most promising of the CH models - one based on satisfying certain similarity relations and the second making use of a scattering phase function that permits only discrete directional changes - can be developed using these approaches. The main idea is to exploit the connection between the space-angle moments of the radiance and the angular moments of the scattering phase function. We compare the results obtained when the two CH models studied are used to simulate an idealized tissue transport problem. The numerical results support our findings based on the theoretical derivations and suggest that CH models should play a useful role in modeling light-tissue interactions
Hydrogen recycle modeling in transport codes
International Nuclear Information System (INIS)
Howe, H.C.
1979-01-01
The hydrogen recycling models now used in Tokamak transport codes are reviewed and the method by which realistic recycling models are being added is discussed. Present models use arbitrary recycle coefficients and therefore do not model the actual recycling processes at the wall. A model for the hydrogen concentration in the wall serves two purposes: (1) it allows a better understanding of the density behavior in present gas puff, pellet, and neutral beam heating experiments; and (2) it allows one to extrapolate to long pulse devices such as EBT, ISX-C and reactors where the walls are observed or expected to saturate. Several wall models are presently being studied for inclusion in transport codes
Validation of comprehensive space radiation transport code
International Nuclear Information System (INIS)
Shinn, J.L.; Simonsen, L.C.; Cucinotta, F.A.
1998-01-01
The HZETRN code has been developed over the past decade to evaluate the local radiation fields within sensitive materials on spacecraft in the space environment. Most of the more important nuclear and atomic processes are now modeled and evaluation within a complex spacecraft geometry with differing material components, including transition effects across boundaries of dissimilar materials, are included. The atomic/nuclear database and transport procedures have received limited validation in laboratory testing with high energy ion beams. The codes have been applied in design of the SAGE-III instrument resulting in material changes to control injurious neutron production, in the study of the Space Shuttle single event upsets, and in validation with space measurements (particle telescopes, tissue equivalent proportional counters, CR-39) on Shuttle and Mir. The present paper reviews the code development and presents recent results in laboratory and space flight validation
Load Balancing of Parallel Monte Carlo Transport Calculations
International Nuclear Information System (INIS)
Procassini, R J; O'Brien, M J; Taylor, J M
2005-01-01
The performance of parallel Monte Carlo transport calculations which use both spatial and particle parallelism is increased by dynamically assigning processors to the most worked domains. Since he particle work load varies over the course of the simulation, this algorithm determines each cycle if dynamic load balancing would speed up the calculation. If load balancing is required, a small number of particle communications are initiated in order to achieve load balance. This method has decreased the parallel run time by more than a factor of three for certain criticality calculations
Dynamic Load Balancing of Parallel Monte Carlo Transport Calculations
International Nuclear Information System (INIS)
O'Brien, M; Taylor, J; Procassini, R
2004-01-01
The performance of parallel Monte Carlo transport calculations which use both spatial and particle parallelism is increased by dynamically assigning processors to the most worked domains. Since the particle work load varies over the course of the simulation, this algorithm determines each cycle if dynamic load balancing would speed up the calculation. If load balancing is required, a small number of particle communications are initiated in order to achieve load balance. This method has decreased the parallel run time by more than a factor of three for certain criticality calculations
Adaptively Learning an Importance Function Using Transport Constrained Monte Carlo
International Nuclear Information System (INIS)
Booth, T.E.
1998-01-01
It is well known that a Monte Carlo estimate can be obtained with zero-variance if an exact importance function for the estimate is known. There are many ways that one might iteratively seek to obtain an ever more exact importance function. This paper describes a method that has obtained ever more exact importance functions that empirically produce an error that is dropping exponentially with computer time. The method described herein constrains the importance function to satisfy the (adjoint) Boltzmann transport equation. This constraint is provided by using the known form of the solution, usually referred to as the Case eigenfunction solution
Reliability analysis of neutron transport simulation using Monte Carlo method
International Nuclear Information System (INIS)
Souza, Bismarck A. de; Borges, Jose C.
1995-01-01
This work presents a statistical and reliability analysis covering data obtained by computer simulation of neutron transport process, using the Monte Carlo method. A general description of the method and its applications is presented. Several simulations, corresponding to slowing down and shielding problems have been accomplished. The influence of the physical dimensions of the materials and of the sample size on the reliability level of results was investigated. The objective was to optimize the sample size, in order to obtain reliable results, optimizing computation time. (author). 5 refs, 8 figs
The KFA-Version of the high-energy transport code HETC and the generalized evaluation code SIMPEL
International Nuclear Information System (INIS)
Cloth, P.; Filges, D.; Sterzenbach, G.; Armstrong, T.W.; Colborn, B.L.
1983-03-01
This document describes the updates that have been made to the high-energy transport code HETC for use in the German spallation-neutron source project SNQ. Performance and purpose of the subsidiary code SIMPEL that has been written for general analysis of the HETC output are also described. In addition means of coupling to low energy transport programs, such as the Monte-Carlo code MORSE is provided. As complete input descriptions for HETC and SIMPEL are given together with a sample problem, this document can serve as a user's manual for these two codes. The document is also an answer to the demand that has been issued by a greater community of HETC users on the ICANS-IV meeting, Oct 20-24 1980, Tsukuba-gun, Japan for a complete description of at least one single version of HETC among the many different versions that exist. (orig.)
Mesh generation and energy group condensation studies for the jaguar deterministic transport code
International Nuclear Information System (INIS)
Kennedy, R. A.; Watson, A. M.; Iwueke, C. I.; Edwards, E. J.
2012-01-01
The deterministic transport code Jaguar is introduced, and the modeling process for Jaguar is demonstrated using a two-dimensional assembly model of the Hoogenboom-Martin Performance Benchmark Problem. This single assembly model is being used to test and analyze optimal modeling methodologies and techniques for Jaguar. This paper focuses on spatial mesh generation and energy condensation techniques. In this summary, the models and processes are defined as well as thermal flux solution comparisons with the Monte Carlo code MC21. (authors)
Mesh generation and energy group condensation studies for the jaguar deterministic transport code
Energy Technology Data Exchange (ETDEWEB)
Kennedy, R. A.; Watson, A. M.; Iwueke, C. I.; Edwards, E. J. [Knolls Atomic Power Laboratory, Bechtel Marine Propulsion Corporation, P.O. Box 1072, Schenectady, NY 12301-1072 (United States)
2012-07-01
The deterministic transport code Jaguar is introduced, and the modeling process for Jaguar is demonstrated using a two-dimensional assembly model of the Hoogenboom-Martin Performance Benchmark Problem. This single assembly model is being used to test and analyze optimal modeling methodologies and techniques for Jaguar. This paper focuses on spatial mesh generation and energy condensation techniques. In this summary, the models and processes are defined as well as thermal flux solution comparisons with the Monte Carlo code MC21. (authors)
Accurate simulation of ionisation chamber response with the Monte Carlo code PENELOPE
International Nuclear Information System (INIS)
Sempau, Josep; Andreo, Pedro
2011-01-01
Ionisation chambers (IC) are routinely used in hospitals for the dosimetry of the photon and electron beams used for radiotherapy treatments. The determination of absorbed dose to water from the absorbed dose to the air filling the cavity requires the introduction of stopping power ratios and perturbation factors, which account for the disturbance caused by the presence of the chamber. Although this may seem a problem readily amenable to Monte Carlo simulation, the fact is that the accurate determination of IC response has been, for various decades, one of the most important challenges of the simulation of electromagnetic showers. The main difficulty stems from the use of condensed history techniques for electron and positron transport. This approach, which involves grouping a large number of interactions into a single artificial event, is known to produce the so-called interface effects when particles travel across surfaces separating different media. These effects can be sizeable when the electron step length is not negligible compared to the size of the region being crossed, as it is the case with the cavity of an IC. The artefact, which becomes apparent when the chamber response shows a marked dependence on the adopted step size, can be palliated with the use of sophisticated electron transport algorithms. These topics are discussed in the context of the transport model implemented in the PENELOPE code. The degree of violation of the Fano theorem for a simple, planar geometry, is used as a measure of the stability of the algorithm with respect to variations of the electron step length, thus assessing the 'quality' of its condensed history scheme. It is shown that, with a suitable choice of transport parameters, PENELOPE simulates IC response with an accuracy of the order of 0.1%.
Accurate simulation of ionization chamber response with the Monte Carlo code PENELOPE
International Nuclear Information System (INIS)
Sempau, Josep
2010-01-01
Full text. Ionization chambers (IC) are routinely used in hospitals for the dosimetry of the photon and electron beams used for radiotherapy treatments. The determination of absorbed dose to water from the absorbed dose to the air filling the cavity requires the introduction of stopping power ratios and perturbation factors, which account for the disturbance caused by the presence of the chamber. Although this may seem a problem readily amenable to Monte Carlo simulation, the fact is that the accurate determination of IC response has been, during the last 20 years, one of the most important challenges of the simulation of electromagnetic showers. The main difficulty stems from the use of condensed history techniques for electron and positron transport. This approach, which involves grouping a large number of interactions into a single artificial event, is known to produce the so-called interface effects when particles travel across surfaces separating different media. These effects are extremely important when the electron step length is not negligible compared to the size of the region being crossed, as it is the case with the cavity of an IC. The artifact, which becomes apparent when the chamber response shows a marked dependence on the adopted step size, can be palliated with the use of sophisticated electron transport algorithms. These topics will be discussed in the context of the transport model implemented in the Penelope code. The degree of violation of the Fano theorem for a simple, planar geometry, will be used as a measure of the stability of the algorithm with respect to variations of the electron step length, thus assessing the 'quality' of its condensed history scheme. It will be shown that, with a suitable choice of transport parameters, Penelope can simulate IC response with an accuracy of the order of 0.1%. (author)
The Monte Carlo code MCBEND - where it is and where it's going
International Nuclear Information System (INIS)
Chukas, S.J.; Miller, P.C.; Power, S.W.
1990-05-01
The Monte Carlo method forms a corner stone to the calculational procedures established in the UK for shielding design and assessment. The emphasis of the work in the shielding area is centred on the Monte Carlo code MCBEND. The work programme in support of the code is broadly directed towards utilisation of new hardware, the development of improved modelling algorithms, the development of new acceleration methods for specific applications and enhancements to user image. This paper summarises the current status of MCBEND and reviews developments carried out over the past two years and planned for the future. (author)
Advances in Monte-Carlo code TRIPOLI-4®’s treatment of the electromagnetic cascade
Directory of Open Access Journals (Sweden)
Mancusi Davide
2018-01-01
Full Text Available TRIPOLI-4® is a Monte-Carlo particle-transport code developed at CEA-Saclay (France that is employed in the domains of nuclear-reactor physics, criticality-safety, shielding/radiation protection and nuclear instrumentation. The goal of this paper is to report on current developments, validation and verification made in TRIPOLI-4 in the electron/positron/photon sector. The new capabilities and improvements concern refinements to the electron transport algorithm, the introduction of a charge-deposition score, the new thick-target bremsstrahlung option, the upgrade of the bremsstrahlung model and the improvement of electron angular straggling at low energy. The importance of each of the developments above is illustrated by comparisons with calculations performed with other codes and with experimental data.
Radiation protection studies for medical particle accelerators using FLUKA Monte Carlo code
International Nuclear Information System (INIS)
Infantino, Angelo; Mostacci, Domiziano; Cicoria, Gianfranco; Lucconi, Giulia; Pancaldi, Davide; Vichi, Sara; Zagni, Federico; Marengo, Mario
2017-01-01
Radiation protection (RP) in the use of medical cyclotrons involves many aspects both in the routine use and for the decommissioning of a site. Guidelines for site planning and installation, as well as for RP assessment, are given in international documents; however, the latter typically offer analytic methods of calculation of shielding and materials activation, in approximate or idealised geometry set-ups. The availability of Monte Carlo (MC) codes with accurate up-to-date libraries for transport and interaction of neutrons and charged particles at energies below 250 MeV, together with the continuously increasing power of modern computers, makes the systematic use of simulations with realistic geometries possible, yielding equipment and site-specific evaluation of the source terms, shielding requirements and all quantities relevant to RP at the same time. In this work, the well-known FLUKA MC code was used to simulate different aspects of RP in the use of biomedical accelerators, particularly for the production of medical radioisotopes. In the context of the Young Professionals Award, held at the IRPA 14 conference, only a part of the complete work is presented. In particular, the simulation of the GE PETtrace cyclotron (16.5 MeV) installed at S. Orsola-Malpighi University Hospital evaluated the effective dose distribution around the equipment; the effective number of neutrons produced per incident proton and their spectral distribution; the activation of the structure of the cyclotron and the vault walls; the activation of the ambient air, in particular the production of "4"1Ar. The simulations were validated, in terms of physical and transport parameters to be used at the energy range of interest, through an extensive measurement campaign of the neutron environmental dose equivalent using a rem-counter and TLD dosemeters. The validated model was then used in the design and the licensing request of a new Positron Emission Tomography facility. (authors)
Report on the Oak Ridge workshop on Monte Carlo codes for relativistic heavy-ion collisions
International Nuclear Information System (INIS)
Awes, T.C.; Sorensen, S.P.
1988-01-01
In order to make detailed predictions for the case of purely hadronic matter, several Monte Carlo codes have been developed to describe relativistic nucleus-nucleus collisions. Although these various models build upon models of hadron-hadron interactions and have been fitted to reproduce hadron-hadron collision data, they have rather different pictures of the underlying hadron collision process and of subsequent particle production. Until now, the different Monte Carlo codes have, in general, been compared to different sets of experimental data, according to which results were readily available to the model builder or which Monte Carlo code was readily available to an experimental group. As a result, it has been difficult to draw firm conclusions about whether the observed deviations between experiments and calculations were due to deficiencies in the particular model, experimental discrepancies, or interesting effects beyond a simple superposition of nucleon-nucleon collisions. For this reason, it was decided that it would be productive to have a structured confrontation between the available experimental data and the many models of high-energy nuclear collisions in a manner in which it could be ensured that the computer codes were run correctly and the experimental acceptances were properly taken into account. With this purpose in mind, a Workshop on Monte Carlo Codes for Relativistic Heavy-Ion Collisions was organized at the Joint Institute for Heavy Ion Research at Oak Ridge National Laboratory from September 12--23, 1988. This paper reviews this workshop. 11 refs., 6 figs
Application of neutron/gamma transport codes for the design of explosive detection systems
International Nuclear Information System (INIS)
Elias, E.; Shayer, Z.
1994-01-01
Applications of neutron and gamma transport codes to the design of nuclear techniques for detecting concealed explosives material are discussed. The methodology of integrating radiation transport computations in the development, optimization and analysis phases of these new technologies is discussed. Transport and Monte Carlo codes are used for proof of concepts, guide the system integration, reduce the extend of experimental program and provide insight into the physical problem involved. The paper concentrates on detection techniques based on thermal and fast neutron interactions in the interrogated object. (authors). 6 refs., 1 tab., 5 figs
Estimation of skyshine dose from turbine building of BWR plant using Monte Carlo code
Energy Technology Data Exchange (ETDEWEB)
Yuji, Nemoto; Toshihisa, Tsukiyama; Shigeki, Nemezawa [Hitachi. Ltd., Saiwai-cho, Hitachi (Japan); Tadashi, Yamasaki; Hidetsugu, Okada [Chubu Electric Power Company, Inc., Odaka-cho, Midori-ku Nagoya (Japan)
2007-07-01
The Monte Carlo N-Particle transport code (MCNP) was adopted to calculate the skyshine dose from the turbine building of a BWR plant for obtaining precise estimations at the site boundary. In MCNP calculation, the equipment and piping arranged on the operating floor of the turbine building were considered and modeled in detail. The inner and outer walls of the turbine building, the shielding materials around the high-pressure turbine, and the piping connected from the moisture separator to the low-pressure turbine were all considered. A three-step study was conducted to estimate the applicability of MCNP code. The first step is confirming the propriety of calculation models. The atmospheric relief diaphragms, which are installed on top of the low-pressure turbine exhaust hood, are not considered in the calculation model. There was little difference between the skyshine dose distributions that were considered when using and not using the atmospheric relief diaphragms. The calculated dose rates agreed well with the measurements taken around the turbine. The second step is estimating the dose rates on the outer roof surface of the turbine building. This calculation was made to confirm the dose distribution of gamma-rays on the turbine roof before being scattered into the air. The calculated dose rates agreed well with the measured data. The third step is making a final confirmation by comparing the calculations and measurements of skyshine dose rates around the turbine building. The source terms of the main steam system are based on the measured activity data of N-16 and C-15. As a conclusion, we were able to calculate reasonable skyshine dose rates by using MCNP code. (authors)
Modeling of the YALINA booster facility by the Monte Carlo code MONK
International Nuclear Information System (INIS)
Talamo, A.; Gohar, Y.; Kondev, F.; Kiyavitskaya, H.; Serafimovich, I.; Bournos, V.; Fokov, Y.; Routkovskaya, C.
2007-01-01
The YALINA-Booster facility has been modeled according to the benchmark specifications defined for the IAEA activity without any geometrical homogenization using the Monte Carlo codes MONK and MCNP/MCNPX/MCB. The MONK model perfectly matches the MCNP one. The computational analyses have been extended through the MCB code, which is an extension of the MCNP code with burnup capability because of its additional feature for analyzing source driven multiplying assemblies. The main neutronics arameters of the YALINA-Booster facility were calculated using these computer codes with different nuclear data libraries based on ENDF/B-VI-0, -6, JEF-2.2, and JEF-3.1.
MKENO-DAR: a direct angular representation Monte Carlo code for criticality safety analysis
International Nuclear Information System (INIS)
Naito, Yoshitaka; Komuro, Yuichi; Tsunoo, Yukiyasu; Nakayama, Mitsuo.
1984-03-01
Improving the Monte Carlo code MULTI-KENO, the MKENO-DAR (Direct Angular Representation) code has been developed for criticality safety analysis in detail. A function was added to MULTI-KENO for representing anisotropic scattering strictly. With this function, the scattering angle of neutron is determined not by the average scattering angle μ-bar of the Pl Legendre polynomial but by the random work operation using probability distribution function produced with the higher order Legendre polynomials. This code is avilable for the FACOM-M380 computer. This report is a computer code manual for MKENO-DAR. (author)
Transport calculation of medium-energy protons and neutrons by Monte Carlo method
International Nuclear Information System (INIS)
Ban, Syuuichi; Hirayama, Hideo; Katoh, Kazuaki.
1978-09-01
A Monte Carlo transport code, ARIES, has been developed for protons and neutrons at medium energy (25 -- 500 MeV). Nuclear data provided by R.G. Alsmiller, Jr. were used for the calculation. To simulate the cascade development in the medium, each generation was represented by a single weighted particle and an average number of emitted particles was used as the weight. Neutron fluxes were stored by the collisions density method. The cutoff energy was set to 25 MeV. Neutrons below the cutoff were stored to be used as the source for the low energy neutron transport calculation upon the discrete ordinates method. Then transport calculations were performed for both low energy neutrons (thermal -- 25 MeV) and secondary gamma-rays. Energy spectra of emitted neutrons were calculated and compared with those of published experimental and calculated results. The agreement was good for the incident particles of energy between 100 and 500 MeV. (author)
SWAT3.1 - the integrated burnup code system driving continuous energy Monte Carlo codes MVP and MCNP
International Nuclear Information System (INIS)
Suyama, Kenya; Mochizuki, Hiroki; Takada, Tomoyuki; Ryufuku, Susumu; Okuno, Hiroshi; Murazaki, Minoru; Ohkubo, Kiyoshi
2009-05-01
Integrated burnup calculation code system SWAT is a system that combines neutronics calculation code SRAC,which is widely used in Japan, and point burnup calculation code ORIGEN2. It has been used to evaluate the composition of the uranium, plutonium, minor actinides and the fission products in the spent nuclear fuel. Based on this idea, the integrated burnup calculation code system SWAT3.1 was developed by combining the continuous energy Monte Carlo code MVP and MCNP, and ORIGEN2. This enables us to treat the arbitrary fuel geometry and to generate the effective cross section data to be used in the burnup calculation with few approximations. This report describes the outline, input data instruction and several examples of the calculation. (author)
Adaptive multilevel splitting for Monte Carlo particle transport
Directory of Open Access Journals (Sweden)
Louvin Henri
2017-01-01
Full Text Available In the Monte Carlo simulation of particle transport, and especially for shielding applications, variance reduction techniques are widely used to help simulate realisations of rare events and reduce the relative errors on the estimated scores for a given computation time. Adaptive Multilevel Splitting (AMS is one of these variance reduction techniques that has recently appeared in the literature. In the present paper, we propose an alternative version of the AMS algorithm, adapted for the first time to the field of particle transport. Within this context, it can be used to build an unbiased estimator of any quantity associated with particle tracks, such as flux, reaction rates or even non-Boltzmann tallies like pulse-height tallies and other spectra. Furthermore, the efficiency of the AMS algorithm is shown not to be very sensitive to variations of its input parameters, which makes it capable of significant variance reduction without requiring extended user effort.
An improved method for storing and retrieving tabulated data in a scalar Monte Carlo code
International Nuclear Information System (INIS)
Hollenbach, D.F.; Reynolds, K.H.; Dodds, H.L.; Landers, N.F.; Petrie, L.M.
1990-01-01
The KENO-Va code is a production-level criticality safety code used to calculate the k eff of a system. The code is stochastic in nature, using a Monte Carlo algorithm to track individual particles one at a time through the system. The advent of computers with vector processors has generated an interest in improving KENO-Va to take advantage of the potential speed-up associated with these new processors. Unfortunately, the original Monte Carlo algorithm and method of storing and retrieving cross-section data is not adaptable to vector processing. This paper discusses an alternate method for storing and retrieving data that not only is readily vectorizable but also improves the efficiency of the current scalar code
Monte-Carlo simulation of complex vapor-transport systems for RIB applications
International Nuclear Information System (INIS)
Zhang, Y.; Alton, G.D.
2005-01-01
In order to minimize decay losses of short-lived radioactive species at ISOL based RIB facilities, effusive-flow particle transit times between target and ion source must be as short as practically achievable. A Monte-Carlo code has been developed for simulating the effusive-flow of neutral particles through vapor-transport systems independent of materials of construction. The code provides average distance traveled and time information associated with the transit of individual particles through a system. It offers a cost effective and accurate means for arriving at vapor-transport system designs. In this report, the code will be described and results obtained by its use in evaluating several prototype vapor-transport systems using specular reflection, cosine and isotropic particle re-emission about the normal to the surface models following adsorption. Simulation results obtained with an isotropic distribution are in close agreement with experimental measurements of the properties of prototype vapor-transport systems fabricated at the Holifield Radioactive Ion Beam Facility
Monte-Carlo code PARJET to simulate e+e--annihilation events via QCD jets
International Nuclear Information System (INIS)
Ritter, S.
1983-01-01
The Monte-Carlo code PARJET simulates exclusive hadronic final states produced in e + e - -annihilation via a virtual photon by two steps: (i) the fragmentation of the original quark-antiquark pair into further partons using results of perturbative QCD in the leading logarithmic approximation (LLA), and (ii) the transition of these parton jets into hadrons on the basis of a chain decay model. Program summary and code description are given. (author)
SAM-CE, Time-Dependent 3-D Neutron Transport, Gamma Transport in Complex Geometry by Monte-Carlo
International Nuclear Information System (INIS)
2003-01-01
1 - Nature of physical problem solved: The SAM-CE system comprises two Monte Carlo codes, SAM-F and SAM-A. SAM-F supersedes the forward Monte Carlo code, SAM-C. SAM-A is an adjoint Monte Carlo code designed to calculate the response due to fields of primary and secondary gamma radiation. The SAM-CE system is a FORTRAN Monte Carlo computer code designed to solve the time-dependent neutron and gamma-ray transport equations in complex three-dimensional geometries. SAM-CE is applicable for forward neutron calculations and for forward as well as adjoint primary gamma-ray calculations. In addition, SAM-CE is applicable for the gamma-ray stage of the coupled neutron-secondary gamma ray problem, which may be solved in either the forward or the adjoint mode. Time-dependent fluxes, and flux functionals such as dose, heating, count rates, etc., are calculated as functions of energy, time and position. Multiple scoring regions are permitted and these may be either finite volume regions or point detectors or both. Other scores of interest, e.g., collision and absorption densities, etc., are also made. 2 - Method of solution: A special feature of SAM-CE is its use of the 'combinatorial geometry' technique which affords the user geometric capabilities exceeding those available with other commonly used geometric packages. All nuclear interaction cross section data (derived from the ENDF for neutrons and from the UNC-format library for gamma-rays) are tabulated in point energy meshes. The energy meshes for neutrons are internally derived, based on built-in convergence criteria and user- supplied tolerances. Tabulated neutron data for each distinct nuclide are in unique and appropriate energy meshes. Both resolved and unresolved resonance parameters from ENDF data files are treated automatically, and extremely precise and detailed descriptions of cross section behaviour is permitted. Such treatment avoids the ambiguities usually associated with multi-group codes, which use flux
NRMC - A GPU code for N-Reverse Monte Carlo modeling of fluids in confined media
Sánchez-Gil, Vicente; Noya, Eva G.; Lomba, Enrique
2017-08-01
NRMC is a parallel code for performing N-Reverse Monte Carlo modeling of fluids in confined media [V. Sánchez-Gil, E.G. Noya, E. Lomba, J. Chem. Phys. 140 (2014) 024504]. This method is an extension of the usual Reverse Monte Carlo method to obtain structural models of confined fluids compatible with experimental diffraction patterns, specifically designed to overcome the problem of slow diffusion that can appear under conditions of tight confinement. Most of the computational time in N-Reverse Monte Carlo modeling is spent in the evaluation of the structure factor for each trial configuration, a calculation that can be easily parallelized. Implementation of the structure factor evaluation in NVIDIA® CUDA so that the code can be run on GPUs leads to a speed up of up to two orders of magnitude.
THREEDANT: A code to perform three-dimensional, neutral particle transport calculations
International Nuclear Information System (INIS)
Alcouffe, R.E.
1994-01-01
The THREEDANT code solves the three-dimensional neutral particle transport equation in its first order, multigroup, discrate ordinate form. The code allows an unlimited number of groups (depending upon the cross section set), angular quadrature up to S-100, and unlimited Pn order again depending upon the cross section set. The code has three options for spatial differencing, diamond with set-to-zero fixup, adaptive weighted diamond, and linear modal. The geometry options are XYZ and RZΘ with a special XYZ option based upon a volume fraction method. This allows objects or bodies of any shape to be modelled as input which gives the code as much geometric description flexibility as the Monte Carlo code MCNP. The transport equation is solved by source iteration accelerated by the DSA method. Both inner and outer iterations are so accelerated. Some results are presented which demonstrate the effectiveness of these techniques. The code is available on several types of computing platforms
Mairani, A; Valente, M; Battistoni, G; Botta, F; Pedroli, G; Ferrari, A; Cremonesi, M; Di Dia, A; Ferrari, M; Fasso, A
2011-01-01
Purpose: The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, FLUKA Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, FLUKA has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one. Methods: FLUKA DPKS have been calculated in both water and compact bone for monoenergetic electrons (10-3 MeV) and for beta emitting isotopes commonly used for therapy ((89)Sr, (90)Y, (131)I, (153)Sm, (177)Lu, (186)Re, and (188)Re). Point isotropic...
The neutrons flux density calculations by Monte Carlo code for the double heterogeneity fuel
International Nuclear Information System (INIS)
Gurevich, M.I.; Brizgalov, V.I.
1994-01-01
This document provides the calculation technique for the fuel elements which consists of the one substance as a matrix and the other substance as the corn embedded in it. This technique can be used in the neutron flux density calculation by the universal Monte Carlo code. The estimation of accuracy is presented too. (authors). 6 refs., 1 fig
Towards a heavy-ion transport capability in the MARS15 Code
International Nuclear Information System (INIS)
Mokhov, N.V.; Gudima, K.K.; Mashnik, S.G.; Rakhno, I.L.; Striganov, S.
2004-01-01
In order to meet the challenges of new accelerator and space projects and further improve modelling of radiation effects in microscopic objects, heavy-ion interaction and transport physics have been recently incorporated into the MARS15 Monte Carlo code. A brief description of new modules is given in comparison with experimental data. The MARS Monte Carlo code is widely used in numerous accelerator, detector, shielding and cosmic ray applications. The needs of the Relativistic Heavy-Ion Collider, Large Hadron Collider, Rare Isotope Accelerator and NASA projects have recently induced adding heavy-ion interaction and transport physics to the MARS15 code. The key modules of the new implementation are described below along with their comparisons to experimental data.
Modeling Monte Carlo of multileaf collimators using the code GEANT4
Energy Technology Data Exchange (ETDEWEB)
Oliveira, Alex C.H.; Lima, Fernando R.A., E-mail: oliveira.ach@yahoo.com, E-mail: falima@cnen.gov.br [Centro Regional de Ciencias Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil); Lima, Luciano S.; Vieira, Jose W., E-mail: lusoulima@yahoo.com.br [Instituto Federal de Educacao, Ciencia e Tecnologia de Pernambuco (IFPE), Recife, PE (Brazil)
2014-07-01
Radiotherapy uses various techniques and equipment for local treatment of cancer. The equipment most often used in radiotherapy to the patient irradiation is linear accelerator (Linac). Among the many algorithms developed for evaluation of dose distributions in radiotherapy planning, the algorithms based on Monte Carlo (MC) methods have proven to be very promising in terms of accuracy by providing more realistic results. The MC simulations for applications in radiotherapy are divided into two parts. In the first, the simulation of the production of the radiation beam by the Linac is performed and then the phase space is generated. The phase space contains information such as energy, position, direction, etc. of millions of particles (photons, electrons, positrons). In the second part the simulation of the transport of particles (sampled phase space) in certain configurations of irradiation field is performed to assess the dose distribution in the patient (or phantom). Accurate modeling of the Linac head is of particular interest in the calculation of dose distributions for intensity modulated radiation therapy (IMRT), where complex intensity distributions are delivered using a multileaf collimator (MLC). The objective of this work is to describe a methodology for modeling MC of MLCs using code Geant4. To exemplify this methodology, the Varian Millennium 120-leaf MLC was modeled, whose physical description is available in BEAMnrc Users Manual (20 11). The dosimetric characteristics (i.e., penumbra, leakage, and tongue-and-groove effect) of this MLC were evaluated. The results agreed with data published in the literature concerning the same MLC. (author)
Analysis of the KANT experiment on beryllium using TRIPOLI-4 Monte Carlo code
International Nuclear Information System (INIS)
Lee, Yi-Kang
2011-01-01
Beryllium is an important material in fusion technology for multiplying neutrons in blankets. However, beryllium nuclear data are differently presented in modern nuclear data evaluations. Recent investigations with the TRIPOLI-4 Monte Carlo simulation of the tritium breeding ratio (TBR) demonstrated that beryllium reaction data are the main source of the calculation uncertainties between ENDF/B-VII.0 and JEFF-3.1. To clarify the calculation uncertainties from data libraries on beryllium, in this study TRIPOLI-4 calculations of the Karlsruhe Neutron Transmission (KANT) experiment have been performed by using ENDF/B-VII.0 and new JEFF-3.1.1 data libraries. The KANT Experiment on beryllium has been used to validate neutron transport codes and nuclear data libraries. An elaborated KANT experiment benchmark has been compiled and published in the NEA/SINBAD database and it has been used as reference in the present work. The neutron multiplication in bulk beryllium assemblies was considered with a central D-T neutron source. Neutron leakage spectra through the 5, 10, and 17 cm thick spherical beryllium shells were calculated and five-group partial leakage multiplications were reported and discussed. In general, improved C/E ratios on neutron leakage multiplications have been obtained. Both ENDF/B-VII.0 and JEFF-3.1.1 beryllium data libraries of TRIPOLI-4 are acceptable now for fusion neutronics calculations.
Energy Technology Data Exchange (ETDEWEB)
Gurevich, M. I.; Oleynik, D. S. [RRC Kurchatov Inst., Kurchatov Sq., 1, 123182, Moscow (Russian Federation); Russkov, A. A.; Voloschenko, A. M. [Keldysh Inst. of Applied Mathematics, Miusskaya Sq., 4, 125047, Moscow (Russian Federation)
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)
penORNL: a parallel Monte Carlo photon and electron transport package using PENELOPE
International Nuclear Information System (INIS)
Bekar, Kursat B.; Miller, Thomas Martin; Patton, Bruce W.; Weber, Charles F.
2015-01-01
The parallel Monte Carlo photon and electron transport code package penORNL was developed at Oak Ridge National Laboratory to enable advanced scanning electron microscope (SEM) simulations on high-performance computing systems. This paper discusses the implementations, capabilities and parallel performance of the new code package. penORNL uses PENELOPE for its physics calculations and provides all available PENELOPE features to the users, as well as some new features including source definitions specifically developed for SEM simulations, a pulse-height tally capability for detailed simulations of gamma and x-ray detectors, and a modified interaction forcing mechanism to enable accurate energy deposition calculations. The parallel performance of penORNL was extensively tested with several model problems, and very good linear parallel scaling was observed with up to 512 processors. penORNL, along with its new features, will be available for SEM simulations upon completion of the new pulse-height tally implementation.
International Nuclear Information System (INIS)
Kim, Jong Woo; Woo, Myeong Hyeon; Kim, Jae Hyun; Kim, Do Hyun; Shin, Chang Ho; Kim, Jong Kyung
2017-01-01
In this study hybrid Monte Carlo/Deterministic method is explained for radiation transport analysis in global system. FW-CADIS methodology construct the weight window parameter and it useful at most global MC calculation. However, Due to the assumption that a particle is scored at a tally, less particles are transported to the periphery of mesh tallies. For compensation this space-dependency, we modified the module in the ADVANTG code to add the proposed method. We solved the simple test problem for comparing with result from FW-CADIS methodology, it was confirmed that a uniform statistical error was secured as intended. In the future, it will be added more practical problems. It might be useful to perform radiation transport analysis using the Hybrid Monte Carlo/Deterministic method in global transport problems.
Icarus: A 2-D Direct Simulation Monte Carlo (DSMC) Code for Multi-Processor Computers
International Nuclear Information System (INIS)
BARTEL, TIMOTHY J.; PLIMPTON, STEVEN J.; GALLIS, MICHAIL A.
2001-01-01
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
Applications of FLUKA Monte Carlo code for nuclear and accelerator physics
Battistoni, Giuseppe; Brugger, Markus; Campanella, Mauro; Carboni, Massimo; Empl, Anton; Fasso, Alberto; Gadioli, Ettore; Cerutti, Francesco; Ferrari, Alfredo; Ferrari, Anna; Lantz, Matthias; Mairani, Andrea; Margiotta, M; Morone, Christina; Muraro, Silvia; Parodi, Katerina; Patera, Vincenzo; Pelliccioni, Maurizio; Pinsky, Lawrence; Ranft, Johannes; Roesler, Stefan; Rollet, Sofia; Sala, Paola R; Santana, Mario; Sarchiapone, Lucia; Sioli, Maximiliano; Smirnov, George; Sommerer, Florian; Theis, Christian; Trovati, Stefania; Villari, R; Vincke, Heinz; Vincke, Helmut; Vlachoudis, Vasilis; Vollaire, Joachim; Zapp, Neil
2011-01-01
FLUKA is a general purpose Monte Carlo code capable of handling all radiation components from thermal energies (for neutrons) or 1keV (for all other particles) to cosmic ray energies and can be applied in many different fields. Presently the code is maintained on Linux. The validity of the physical models implemented in FLUKA has been benchmarked against a variety of experimental data over a wide energy range, from accelerator data to cosmic ray showers in the Earth atmosphere. FLUKA is widely used for studies related both to basic research and to applications in particle accelerators, radiation protection and dosimetry, including the specific issue of radiation damage in space missions, radiobiology (including radiotherapy) and cosmic ray calculations. After a short description of the main features that make FLUKA valuable for these topics, the present paper summarizes some of the recent applications of the FLUKA Monte Carlo code in the nuclear as well high energy physics. In particular it addresses such top...
ITS, TIGER System of Coupled Electron Photon Transport by Monte-Carlo
International Nuclear Information System (INIS)
Halbleib, J.A.; Mehlhorn, T.A.; Young, M.F.
1996-01-01
1 - Description of program or function: ITS permits a state-of-the-art Monte Carlo solution of linear time-integrated coupled electron/ photon radiation transport problems with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. 2 - Method of solution: Through a machine-portable utility that emulates the basic features of the CDC UPDATE processor, the user selects one of eight codes for running on a machine of one of four (at least) major vendors. With the ITS-3.0 release the PSR-0245/UPEML package is included to perform these functions. The ease with which this utility is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is maximized by employing the best available cross sections and sampling distributions, and the most complete physical model for describing the production and transport of the electron/ photon cascade from 1.0 GeV down to 1.0 keV. Flexibility of construction permits the codes to be tailored to specific applications and the capabilities of the codes to be extended to more complex applications through update procedures. 3 - Restrictions on the complexity of the problem: - Restrictions and/or limitations for ITS depend upon the local operating system
bhlight: GENERAL RELATIVISTIC RADIATION MAGNETOHYDRODYNAMICS WITH MONTE CARLO TRANSPORT
International Nuclear Information System (INIS)
Ryan, B. R.; Gammie, C. F.; Dolence, J. C.
2015-01-01
We present bhlight, a numerical scheme for solving the equations of general relativistic radiation magnetohydrodynamics using a direct Monte Carlo solution of the frequency-dependent radiative transport equation. bhlight is designed to evolve black hole accretion flows at intermediate accretion rate, in the regime between the classical radiatively efficient disk and the radiatively inefficient accretion flow (RIAF), in which global radiative effects play a sub-dominant but non-negligible role in disk dynamics. We describe the governing equations, numerical method, idiosyncrasies of our implementation, and a suite of test and convergence results. We also describe example applications to radiative Bondi accretion and to a slowly accreting Kerr black hole in axisymmetry
International Nuclear Information System (INIS)
Bécares, V.; Pérez-Martín, S.; Vázquez-Antolín, M.; Villamarín, D.; Martín-Fuertes, F.; González-Romero, E.M.; Merino, I.
2014-01-01
Highlights: • Review of several Monte Carlo effective delayed neutron fraction calculation methods. • These methods have been implemented with the Monte Carlo code MCNPX. • They have been benchmarked against against some critical and subcritical systems. • Several nuclear data libraries have been used. - Abstract: The calculation of the effective delayed neutron fraction, β eff , with Monte Carlo codes is a complex task due to the requirement of properly considering the adjoint weighting of delayed neutrons. Nevertheless, several techniques have been proposed to circumvent this difficulty and obtain accurate Monte Carlo results for β eff without the need of explicitly determining the adjoint flux. In this paper, we make a review of some of these techniques; namely we have analyzed two variants of what we call the k-eigenvalue technique and other techniques based on different interpretations of the physical meaning of the adjoint weighting. To test the validity of all these techniques we have implemented them with the MCNPX code and we have benchmarked them against a range of critical and subcritical systems for which either experimental or deterministic values of β eff are available. Furthermore, several nuclear data libraries have been used in order to assess the impact of the uncertainty in nuclear data in the calculated value of β eff
Error reduction techniques for Monte Carlo neutron transport calculations
International Nuclear Information System (INIS)
Ju, J.H.W.
1981-01-01
Monte Carlo methods have been widely applied to problems in nuclear physics, mathematical reliability, communication theory, and other areas. The work in this thesis is developed mainly with neutron transport applications in mind. For nuclear reactor and many other applications, random walk processes have been used to estimate multi-dimensional integrals and obtain information about the solution of integral equations. When the analysis is statistically based such calculations are often costly, and the development of efficient estimation techniques plays a critical role in these applications. All of the error reduction techniques developed in this work are applied to model problems. It is found that the nearly optimal parameters selected by the analytic method for use with GWAN estimator are nearly identical to parameters selected by the multistage method. Modified path length estimation (based on the path length importance measure) leads to excellent error reduction in all model problems examined. Finally, it should be pointed out that techniques used for neutron transport problems may be transferred easily to other application areas which are based on random walk processes. The transport problems studied in this dissertation provide exceptionally severe tests of the error reduction potential of any sampling procedure. It is therefore expected that the methods of this dissertation will prove useful in many other application areas
Monte Carlo methods for flux expansion solutions of transport problems
International Nuclear Information System (INIS)
Spanier, J.
1999-01-01
Adaptive Monte Carlo methods, based on the use of either correlated sampling or importance sampling, to obtain global solutions to certain transport problems have recently been described. The resulting learning algorithms are capable of achieving geometric convergence when applied to the estimation of a finite number of coefficients in a flux expansion representation of the global solution. However, because of the nonphysical nature of the random walk simulations needed to perform importance sampling, conventional transport estimators and source sampling techniques require modification to be used successfully in conjunction with such flux expansion methods. It is shown how these problems can be overcome. First, the traditional path length estimators in wide use in particle transport simulations are generalized to include rather general detector functions (which, in this application, are the individual basis functions chosen for the flus expansion). Second, it is shown how to sample from the signed probabilities that arise as source density functions in these applications, without destroying the zero variance property needed to ensure geometric convergence to zero error
OGRE, Monte-Carlo System for Gamma Transport Problems
International Nuclear Information System (INIS)
1984-01-01
1 - Nature of physical problem solved: The OGRE programme system was designed to calculate, by Monte Carlo methods, any quantity related to gamma-ray transport. The system is represented by two examples - OGRE-P1 and OGRE-G. The OGRE-P1 programme is a simple prototype which calculates dose rate on one side of a slab due to a plane source on the other side. The OGRE-G programme, a prototype of a programme utilizing a general-geometry routine, calculates dose rate at arbitrary points. A very general source description in OGRE-G may be employed by reading a tape prepared by the user. 2 - Method of solution: Case histories of gamma rays in the prescribed geometry are generated and analyzed to produce averages of any desired quantity which, in the case of the prototypes, are gamma-ray dose rates. The system is designed to achieve generality by ease of modification. No importance sampling is built into the prototypes, a very general geometry subroutine permits the treatment of complicated geometries. This is essentially the same routine used in the O5R neutron transport system. Boundaries may be either planes or quadratic surfaces, arbitrarily oriented and intersecting in arbitrary fashion. Cross section data is prepared by the auxiliary master cross section programme XSECT which may be used to originate, update, or edit the master cross section tape. The master cross section tape is utilized in the OGRE programmes to produce detailed tables of macroscopic cross sections which are used during the Monte Carlo calculations. 3 - Restrictions on the complexity of the problem: Maximum cross-section array information may be estimated by a given formula for a specific problem. The number of regions must be less than or equal to 50
MONTEBURNS 2.0: An Automated, Multi-Step Monte Carlo Burnup Code System
International Nuclear Information System (INIS)
2007-01-01
A - Description of program or function: 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, nor from the NEA Databank. 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. The following changes have been made: 1) An increase in the number of removal group information that must be provided for each material in each step in the feed input file. 2) The capability to use CINDER90 instead of ORIGEN2.1 as the depletion/decay part of the code. 3) ORIGEN2.2 can also be used instead of ORIGEN2.1 in Monteburns. 4) The correction of including the capture cross sections to metastable as well as ground states if applicable for an isotope (i.e. Am-241 and Am-243 in particular). 5) The ability to use a MCNP input file that has a title card starting with 'm' (this was a bug in the first version of Monteburns). 6) A decrease in run time for cases involving decay-only steps (power of 0.0). Monteburns does not run MCNP to calculate cross sections for a step unless it is an irradiation step. 7) The ability to change the cross section libraries used each step. If different cross section libraries are desired for multiple steps. 8
Foucart, Francois
2018-04-01
General relativistic radiation hydrodynamic simulations are necessary to accurately model a number of astrophysical systems involving black holes and neutron stars. Photon transport plays a crucial role in radiatively dominated accretion discs, while neutrino transport is critical to core-collapse supernovae and to the modelling of electromagnetic transients and nucleosynthesis in neutron star mergers. However, evolving the full Boltzmann equations of radiative transport is extremely expensive. Here, we describe the implementation in the general relativistic SPEC code of a cheaper radiation hydrodynamic method that theoretically converges to a solution of Boltzmann's equation in the limit of infinite numerical resources. The algorithm is based on a grey two-moment scheme, in which we evolve the energy density and momentum density of the radiation. Two-moment schemes require a closure that fills in missing information about the energy spectrum and higher order moments of the radiation. Instead of the approximate analytical closure currently used in core-collapse and merger simulations, we complement the two-moment scheme with a low-accuracy Monte Carlo evolution. The Monte Carlo results can provide any or all of the missing information in the evolution of the moments, as desired by the user. As a first test of our methods, we study a set of idealized problems demonstrating that our algorithm performs significantly better than existing analytical closures. We also discuss the current limitations of our method, in particular open questions regarding the stability of the fully coupled scheme.
TMCC: a transient three-dimensional neutron transport code by the direct simulation method - 222
International Nuclear Information System (INIS)
Shen, H.; Li, Z.; Wang, K.; Yu, G.
2010-01-01
A direct simulation method (DSM) is applied to solve the transient three-dimensional neutron transport problems. DSM is based on the Monte Carlo method, and can be considered as an application of the Monte Carlo method in the specific type of problems. In this work, the transient neutronics problem is solved by simulating the dynamic behaviors of neutrons and precursors of delayed neutrons during the transient process. DSM gets rid of various approximations which are always necessary to other methods, so it is precise and flexible in the requirement of geometric configurations, material compositions and energy spectrum. In this paper, the theory of DSM is introduced first, and the numerical results obtained with the new transient analysis code, named TMCC (Transient Monte Carlo Code), are presented. (authors)
ICF target 2D modeling using Monte Carlo SNB electron thermal transport in DRACO
Chenhall, Jeffrey; Cao, Duc; Moses, Gregory
2016-10-01
The iSNB (implicit Schurtz Nicolai Busquet multigroup diffusion electron thermal transport method is adapted into a Monte Carlo (MC) transport method to better model angular and long mean free path non-local effects. The MC model was first implemented in the 1D LILAC code to verify consistency with the iSNB model. Implementation of the MC SNB model in the 2D DRACO code enables higher fidelity non-local thermal transport modeling in 2D implosions such as polar drive experiments on NIF. The final step is to optimize the MC model by hybridizing it with a MC version of the iSNB diffusion method. The hybrid method will combine the efficiency of a diffusion method in intermediate mean free path regions with the accuracy of a transport method in long mean free path regions allowing for improved computational efficiency while maintaining accuracy. Work to date on the method will be presented. This work was supported by Sandia National Laboratories and the Univ. of Rochester Laboratory for Laser Energetics.
Energy Technology Data Exchange (ETDEWEB)
Kurosu, K [Department of Radiation Oncology, Osaka University Graduate School of Medicine, Osaka (Japan); Department of Medical Physics ' Engineering, Osaka University Graduate School of Medicine, Osaka (Japan); Takashina, M; Koizumi, M [Department of Medical Physics ' Engineering, Osaka University Graduate School of Medicine, Osaka (Japan); Das, I; Moskvin, V [Department of Radiation Oncology, Indiana University School of Medicine, Indianapolis, IN (United States)
2014-06-01
Purpose: Monte Carlo codes are becoming important tools for proton beam dosimetry. However, the relationships between the customizing parameters and percentage depth dose (PDD) of GATE and PHITS codes have not been reported which are studied for PDD and proton range compared to the FLUKA code and the experimental data. Methods: The beam delivery system of the Indiana University Health Proton Therapy Center was modeled for the uniform scanning beam in FLUKA and transferred identically into GATE and PHITS. This computational model was built from the blue print and validated with the commissioning data. Three parameters evaluated are the maximum step size, cut off energy and physical and transport model. The dependence of the PDDs on the customizing parameters was compared with the published results of previous studies. Results: The optimal parameters for the simulation of the whole beam delivery system were defined by referring to the calculation results obtained with each parameter. Although the PDDs from FLUKA and the experimental data show a good agreement, those of GATE and PHITS obtained with our optimal parameters show a minor discrepancy. The measured proton range R90 was 269.37 mm, compared to the calculated range of 269.63 mm, 268.96 mm, and 270.85 mm with FLUKA, GATE and PHITS, respectively. Conclusion: We evaluated the dependence of the results for PDDs obtained with GATE and PHITS Monte Carlo generalpurpose codes on the customizing parameters by using the whole computational model of the treatment nozzle. The optimal parameters for the simulation were then defined by referring to the calculation results. The physical model, particle transport mechanics and the different geometrybased descriptions need accurate customization in three simulation codes to agree with experimental data for artifact-free Monte Carlo simulation. This study was supported by Grants-in Aid for Cancer Research (H22-3rd Term Cancer Control-General-043) from the Ministry of Health
International Nuclear Information System (INIS)
Charles A. Wemple; Joshua J. Cogliati
2005-01-01
A univel geometry, neutral particle Monte Carlo transport code, written entirely in the Java programming language, is under development for medical radiotherapy applications. The code uses ENDF-VI based continuous energy cross section data in a flexible XML format. Full neutron-photon coupling, including detailed photon production and photonuclear reactions, is included. Charged particle equilibrium is assumed within the patient model so that detailed transport of electrons produced by photon interactions may be neglected. External beam and internal distributed source descriptions for mixed neutron-photon sources are allowed. Flux and dose tallies are performed on a univel basis. A four-tap, shift-register-sequence random number generator is used. Initial verification and validation testing of the basic neutron transport routines is underway. The searchlight problem was chosen as a suitable first application because of the simplicity of the physical model. Results show excellent agreement with analytic solutions. Computation times for similar numbers of histories are comparable to other neutron MC codes written in C and FORTRAN
Energy Technology Data Exchange (ETDEWEB)
Ford, R.L.; Nelson, W.R.
1978-06-01
A code to simulate almost any electron--photon transport problem conceivable is described. The report begins with a lengthy historical introduction and a description of the shower generation process. Then the detailed physics of the shower processes and the methods used to simulate them are presented. Ideas of sampling theory, transport techniques, particle interactions in general, and programing details are discussed. Next, EGS calculations and various experiments and other Monte Carlo results are compared. The remainder of the report consists of user manuals for EGS, PEGS, and TESTSR codes; options, input specifications, and typical output are included. 38 figures, 12 tables. (RWR)
G4-STORK: A Geant4-based Monte Carlo reactor kinetics simulation code
International Nuclear Information System (INIS)
Russell, Liam; Buijs, Adriaan; Jonkmans, Guy
2014-01-01
Highlights: • G4-STORK is a new, time-dependent, Monte Carlo code for reactor physics applications. • G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. • G4-STORK was designed to simulate short-term fluctuations in reactor cores. • G4-STORK is well suited for simulating sub- and supercritical assemblies. • G4-STORK was verified through comparisons with DRAGON and MCNP. - Abstract: In this paper we introduce G4-STORK (Geant4 STOchastic Reactor Kinetics), a new, time-dependent, Monte Carlo particle tracking code for reactor physics applications. G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. The toolkit provides the fundamental physics models and particle tracking algorithms that track each particle in space and time. It is a framework for further development (e.g. for projects such as G4-STORK). G4-STORK derives reactor physics parameters (e.g. k eff ) from the continuous evolution of a population of neutrons in space and time in the given simulation geometry. In this paper we detail the major additions to the Geant4 toolkit that were necessary to create G4-STORK. These include a renormalization process that maintains a manageable number of neutrons in the simulation even in very sub- or supercritical systems, scoring processes (e.g. recording fission locations, total neutrons produced and lost, etc.) that allow G4-STORK to calculate the reactor physics parameters, and dynamic simulation geometries that can change over the course of simulation to illicit reactor kinetics responses (e.g. fuel temperature reactivity feedback). The additions are verified through simple simulations and code-to-code comparisons with established reactor physics codes such as DRAGON and MCNP. Additionally, G4-STORK was developed to run a single simulation in parallel over many processors using MPI (Message Passing Interface) pipes
Monte Carlo simulation of radioactive contaminant transport in unsaturated porous media
International Nuclear Information System (INIS)
Giacobbo, F.; Patelli, E.; Zio, E.
2005-01-01
In the current proposed solutions of radioactive waste repositories, the protective function against the radionuclide water-driven transport back to the biosphere is to be provided by an integrated system of artificial and natural geologic barriers. The complexity of the transport process in the barriers' heterogeneous media forces approximations to the classical analytical-numerical models, thus reducing their adherence to reality. In an attempt to overcome these difficulties, in the present paper we adopt a Monte Carlo simulation approach, previously developed on the basis of the Kolmogorov and Dmitriev theory of branching stochastic processes. The approach is here extended for describing transport through unsaturated porous media under unsteady flow conditions. This generalization entails the determination of the functional dependence of the parameters of the proposed transport model from the water content, which changes in space and time during the water infiltration process. The approach is verified with respect to a case of non-reactive transport under transient unsaturated field conditions by a comparison with a standard code based on the classical advection-dispersion equations. An application regarding linear reactive transport is then presented. (authors)
The development of depletion program coupled with Monte Carlo computer code
International Nuclear Information System (INIS)
Nguyen Kien Cuong; Huynh Ton Nghiem; Vuong Huu Tan
2015-01-01
The paper presents the development of depletion code for light water reactor coupled with MCNP5 code called the MCDL code (Monte Carlo Depletion for Light Water Reactor). The first order differential depletion system equations of 21 actinide isotopes and 50 fission product isotopes are solved by the Radau IIA Implicit Runge Kutta (IRK) method after receiving neutron flux, reaction rates in one group energy and multiplication factors for fuel pin, fuel assembly or whole reactor core from the calculation results of the MCNP5 code. The calculation for beryllium poisoning and cooling time is also integrated in the code. To verify and validate the MCDL code, high enriched uranium (HEU) and low enriched uranium (LEU) fuel assemblies VVR-M2 types and 89 fresh HEU fuel assemblies, 92 LEU fresh fuel assemblies cores of the Dalat Nuclear Research Reactor (DNRR) have been investigated and compared with the results calculated by the SRAC code and the MCNP R EBUS linkage system code. The results show good agreement between calculated data of the MCDL code and reference codes. (author)
International Nuclear Information System (INIS)
Brown, Forrest B.; Martin, William R.
2001-01-01
We have investigated Monte Carlo schemes for analyzing particle transport through media with exponentially varying time-dependent cross sections. For such media, the cross sections are represented in the form Σ(t) = Σ 0 e -at (1) or equivalently as Σ(x) = Σ 0 e -bx (2) where b = av and v is the particle speed. For the following discussion, the parameters a and b may be either positive, for exponentially decreasing cross sections, or negative, for exponentially increasing cross sections. For most time-dependent Monte Carlo applications, the time and spatial variations of the cross-section data are handled by means of a stepwise procedure, holding the cross sections constant for each region over a small time interval Δt, performing the Monte Carlo random walk over the interval Δt, updating the cross sections, and then repeating for a series of time intervals. Continuously varying spatial- or time-dependent cross sections can be treated in a rigorous Monte Carlo fashion using delta-tracking, but inefficiencies may arise if the range of cross-section variation is large. In this paper, we present a new method for sampling collision distances directly for cross sections that vary exponentially in space or time. The method is exact and efficient and has direct application to Monte Carlo radiation transport methods. To verify that the probability density function (PDF) is correct and that the random-sampling procedure yields correct results, numerical experiments were performed using a one-dimensional Monte Carlo code. The physical problem consisted of a beam source impinging on a purely absorbing infinite slab, with a slab thickness of 1 cm and Σ 0 = 1 cm -1 . Monte Carlo calculations with 10 000 particles were run for a range of the exponential parameter b from -5 to +20 cm -1 . Two separate Monte Carlo calculations were run for each choice of b, a continuously varying case using the random-sampling procedures described earlier, and a 'conventional' case where the
International Nuclear Information System (INIS)
Santa Cruz, G.A.
1998-01-01
Full text: A charged particles transport Monte Carlo code, specially designed for the boron neutron capture therapy microdosimetry study was developed. The code allows the use of real tri dimensional problem geometry, using serial microscopy slides from a biological substrate where the 10 B(n, Alpha) 7 Li, 14 N(n,p) 14 C reactions and events can occur. The spatial distribution of sources ( 10 B, 14 N concentrations), regions of interest (where the energy deposition, linear energy transfer and other parameters will be calculated) and other zones (without boron) are obtained from the images. The code is in the benchmarking stage, using geometrically simple cases and experimental data obtained from microdosimetric spectra from TEPC (Tissue Equivalent Proportional Counters) doped with 10 B. It allows to obtain LET spectra discriminated by event classes, chord-length distributions, dose and frequency mean values and visualizations of the spatial energy deposition. A similar version of the code uses bidimensional images from a tissue sample containing a great number of cellular structures. An equivalence between the microdosimetry of a bidimensional case and a tri dimensional one can be done. If the real distribution of 10 B is known, for example by high resolution alpha-track autoradiography, the code can use this information explicitly. (author) [es
Energy Technology Data Exchange (ETDEWEB)
Quinteiro, Guillermo F; Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Reactores y Centrales Nucleares
2000-07-01
It was carried out for the first time, a model of the Argentine RA-1 reactor using the MCNP Monte Carlo code. This model was validated using data for experimental neutron and gamma measurements at different energy ranges and locations. In addition, the resulting fluxes were compared with the data obtained using a 3D diffusion code. (author)
Energy Technology Data Exchange (ETDEWEB)
Liu, T; Lin, H; Xu, X [Rensselaer Polytechnic Institute, Troy, NY (United States); Stabin, M [Vanderbilt Univ Medical Ctr, Nashville, TN (United States)
2015-06-15
Purpose: To develop a nuclear medicine dosimetry module for the GPU-based Monte Carlo code ARCHER. Methods: We have developed a nuclear medicine dosimetry module for the fast Monte Carlo code ARCHER. The coupled electron-photon Monte Carlo transport kernel included in ARCHER is built upon the Dose Planning Method code (DPM). The developed module manages the radioactive decay simulation by consecutively tracking several types of radiation on a per disintegration basis using the statistical sampling method. Optimization techniques such as persistent threads and prefetching are studied and implemented. The developed module is verified against the VIDA code, which is based on Geant4 toolkit and has previously been verified against OLINDA/EXM. A voxelized geometry is used in the preliminary test: a sphere made of ICRP soft tissue is surrounded by a box filled with water. Uniform activity distribution of I-131 is assumed in the sphere. Results: The self-absorption dose factors (mGy/MBqs) of the sphere with varying diameters are calculated by ARCHER and VIDA respectively. ARCHER’s result is in agreement with VIDA’s that are obtained from a previous publication. VIDA takes hours of CPU time to finish the computation, while it takes ARCHER 4.31 seconds for the 12.4-cm uniform activity sphere case. For a fairer CPU-GPU comparison, more effort will be made to eliminate the algorithmic differences. Conclusion: The coupled electron-photon Monte Carlo code ARCHER has been extended to radioactive decay simulation for nuclear medicine dosimetry. The developed code exhibits good performance in our preliminary test. The GPU-based Monte Carlo code is developed with grant support from the National Institute of Biomedical Imaging and Bioengineering through an R01 grant (R01EB015478)
Comparison of MCB and MONTEBURNS Monte Carlo burnup codes on a one-pass deep burn
International Nuclear Information System (INIS)
Talamo, Alberto; Ji, Wei; Cetnar, Jerzy; Gudowski, Waclaw
2006-01-01
Numerical applications implemented on the Monte Carlo method have developed in line with the increase of computer power; nowadays, in the field of nuclear reactor physics, it is possible to perform burnup simulations in a detailed 3D geometry and a continuous energy description by the Monte Carlo method; moreover, the required computing time can be abundantly reduced by taking advantage of a computer cluster. In this paper we focused on comparing the results of the two major Monte Carlo burnup codes, MONTEBURNS and MCB, when they share the same MCNP geometry, nuclear data library, core thermal power, and they apply the same refueling and shuffling schedule. While simulating a total operation time of the Gas Turbine-Modular Helium Reactor of 2100 effective full power days and a one-pass deep burn in-core fuel management schedule, we have found that the two Monte Carlo codes produce very similar results both on the criticality value of the core and the transmutation of the key actinides
Comparison of MCB and MONTEBURNS Monte Carlo burnup codes on a one-pass deep burn
Energy Technology Data Exchange (ETDEWEB)
Talamo, Alberto [Royal Institute of Technology (KTH), Roslagstullsbacken 21, Stockholm S-10691 (Sweden)]. E-mail: alby@anl.gov; Ji, Wei [University of Michigan, Bonisteel Boulevard 2355, Ann Arbor, MI 48109-2104 (United States); Cetnar, Jerzy [AGH-University of Science and Technology, Al. Mickiewicza 30 Cracow (Poland); Gudowski, Waclaw [Royal Institute of Technology (KTH), Roslagstullsbacken 21, Stockholm S-10691 (Sweden)
2006-09-15
Numerical applications implemented on the Monte Carlo method have developed in line with the increase of computer power; nowadays, in the field of nuclear reactor physics, it is possible to perform burnup simulations in a detailed 3D geometry and a continuous energy description by the Monte Carlo method; moreover, the required computing time can be abundantly reduced by taking advantage of a computer cluster. In this paper we focused on comparing the results of the two major Monte Carlo burnup codes, MONTEBURNS and MCB, when they share the same MCNP geometry, nuclear data library, core thermal power, and they apply the same refueling and shuffling schedule. While simulating a total operation time of the Gas Turbine-Modular Helium Reactor of 2100 effective full power days and a one-pass deep burn in-core fuel management schedule, we have found that the two Monte Carlo codes produce very similar results both on the criticality value of the core and the transmutation of the key actinides.
International Nuclear Information System (INIS)
Karriem, Z.; Ivanov, K.; Zamonsky, O.
2011-01-01
This paper presents work that has been performed to develop an integrated Monte Carlo- Deterministic transport methodology in which the two methods make use of exactly the same general geometry and multigroup nuclear data. The envisioned application of this methodology is in reactor lattice physics methods development and shielding calculations. The methodology will be based on the Method of Long Characteristics (MOC) and the Monte Carlo N-Particle Transport code MCNP5. Important initial developments pertaining to ray tracing and the development of an MOC flux solver for the proposed methodology are described. Results showing the viability of the methodology are presented for two 2-D general geometry transport problems. The essential developments presented is the use of MCNP as geometry construction and ray tracing tool for the MOC, verification of the ray tracing indexing scheme that was developed to represent the MCNP geometry in the MOC and the verification of the prototype 2-D MOC flux solver. (author)
Setup of HDRK-Man voxel model in Geant4 Monte Carlo code
Energy Technology Data Exchange (ETDEWEB)
Jeong, Jong Hwi; Cho, Sung Koo; Kim, Chan Hyeong [Hanyang Univ., Seoul (Korea, Republic of); Choi, Sang Hyoun [Inha Univ., Incheon (Korea, Republic of); Cho, Kun Woo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2008-10-15
Many different voxel models, developed using tomographic images of human body, are used in various fields including both ionizing and non-ionizing radiation fields. Recently a high-quality voxel model/ named HDRK-Man, was constructed at Hanyang University and used to calculate the dose conversion coefficients (DCC) values for external photon and neutron beams using the MCNPX Monte Carlo code. The objective of the present study is to set up the HDRK-Man model in Geant4 in order to use it in more advanced calculations such as 4-D Monte Carlo simulations and space dosimetry studies involving very high energy particles. To that end, the HDRK-Man was ported to Geant4 and used to calculate the DCC values for external photon beams. The calculated values were then compared with the results of the MCNPX code. In addition, a computational Linux cluster was built to improve the computing speed in Geant4.
Evaluation of CASMO-3 and HELIOS for Fuel Assembly Analysis from Monte Carlo Code
Energy Technology Data Exchange (ETDEWEB)
Shim, Hyung Jin; Song, Jae Seung; Lee, Chung Chan
2007-05-15
This report presents a study comparing deterministic lattice physics calculations with Monte Carlo calculations for LWR fuel pin and assembly problems. The study has focused on comparing results from the lattice physics code CASMO-3 and HELIOS against those from the continuous-energy Monte Carlo code McCARD. The comparisons include k{sub inf}, isotopic number densities, and pin power distributions. The CASMO-3 and HELIOS calculations for the k{sub inf}'s of the LWR fuel pin problems show good agreement with McCARD within 956pcm and 658pcm, respectively. For the assembly problems with Gadolinia burnable poison rods, the largest difference between the k{sub inf}'s is 1463pcm with CASMO-3 and 1141pcm with HELIOS. RMS errors for the pin power distributions of CASMO-3 and HELIOS are within 1.3% and 1.5%, respectively.
Energy Technology Data Exchange (ETDEWEB)
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.
Probability-neighbor method of accelerating geometry treatment in reactor Monte Carlo code RMC
International Nuclear Information System (INIS)
She, Ding; Li, Zeguang; Xu, Qi; Wang, Kan; Yu, Ganglin
2011-01-01
Probability neighbor method (PNM) is proposed in this paper to accelerate geometry treatment of Monte Carlo (MC) simulation and validated in self-developed reactor Monte Carlo code RMC. During MC simulation by either ray-tracking or delta-tracking method, large amounts of time are spent in finding out which cell one particle is located in. The traditional way is to search cells one by one with certain sequence defined previously. However, this procedure becomes very time-consuming when the system contains a large number of cells. Considering that particles have different probability to enter different cells, PNM method optimizes the searching sequence, i.e., the cells with larger probability are searched preferentially. The PNM method is implemented in RMC code and the numerical results show that the considerable time of geometry treatment in MC calculation for complicated systems is saved, especially effective in delta-tracking simulation. (author)
Energy Technology Data Exchange (ETDEWEB)
Malatara, G; Kappas, K [Medical Physics Department, Faculty of Medicine, University of Patras, 265 00 Patras (Greece); Sphiris, N [Ethnodata S.A., Athens (Greece)
1994-12-31
In this work, a Monte Carlo EGS4 code was used to simulate radiation transport through linear accelerators to produce and score energy spectra and angular distributions of 6, 12, 15 and 25 MeV bremsstrahlung photons exiting from different accelerator treatment heads. The energy spectra was used as input for a convolution method program to calculate the tissue-maximum ratio in water. 100.000 histories are recorded in the scoring plane for each simulation. The validity of the Monte Carlo simulation and the precision to the calculated spectra have been verified experimentally and were in good agreement. We believe that the accurate simulation of the different components of the linear accelerator head is very important for the precision of the results. The results of the Monte Carlo and the Convolution Method can be compared with experimental data for verification and they are powerful and practical tools to generate accurate spectra and dosimetric data. (authors). 10 refs,5 figs, 2 tabs.
International Nuclear Information System (INIS)
De Matteis, A.
1987-01-01
This report describes the fully automatic linkage between the finite difference, two-dimensional code EDGE2D, based on the classical Braginskii partial differential equations of ion transport, and the Monte Carlo code NIMBUS, which solves the integral form of the stationary, linear Boltzmann equation for neutral transport in a plasma. The coupling has been performed for the real poloidal geometry of JET with two belt-limiters and real magnetic configurations with or without a single-null point. The new integrated system starts from the magnetic geometry computed by predictive or interpretative equilibrium codes and yields the plasma and neutrals characteristics in the edge
A quick and easy improvement of Monte Carlo codes for simulation
Lebrere, A.; Talhi, R.; Tripathy, M.; Pyée, M.
The simulation of trials of independent random variables of given distribution is a critical element of running Monte-Carlo codes. This is usually performed by using pseudo-random number generators (and in most cases linearcongruential ones). We present here an alternative way to generate sequences with given statistical properties. This sequences are purely deterministic and are given by closed formulae, and can give in some cases better results than classical generators.
The codes WAV3BDY and WAV4BDY and the variational Monte Carlo method
International Nuclear Information System (INIS)
Schiavilla, R.
1987-01-01
A description of the codes WAV3BDY and WAV4BDY, which generate the variational ground state wave functions of the A=3 and 4 nuclei, is given, followed by a discussion of the Monte Carlo integration technique, which is used to calculate expectation values and transition amplitudes of operators, and for whose implementation WAV3BDY and WAV4BDY are well suited
Modelling the IRSN's radio-photo-luminescent dosimeter using the MCPNX Monte Carlo code
International Nuclear Information System (INIS)
Hocine, N.; Donadille, L.; Huet, Ch.; Itie, Ch.
2010-01-01
The authors report the modelling of the new radio-photo-luminescent (RPL) dosimeter of the IRSN using the MCPNX Monte Carlo code. The Hp(10) and Hp(0, 07) dose equivalents are computed for different irradiation configurations involving photonic beams (gamma and X) defined according to the ISO 4037-1 standard. Results are compared to experimental measurements performed on the RPL dosimeter. The agreement is good and the model is thus validated
Liu, Tianyu; Wolfe, Noah; Lin, Hui; Zieb, Kris; Ji, Wei; Caracappa, Peter; Carothers, Christopher; Xu, X. George
2017-09-01
This paper contains two parts revolving around Monte Carlo transport simulation on Intel Many Integrated Core coprocessors (MIC, also known as Xeon Phi). (1) MCNP 6.1 was recompiled into multithreading (OpenMP) and multiprocessing (MPI) forms respectively without modification to the source code. The new codes were tested on a 60-core 5110P MIC. The test case was FS7ONNi, a radiation shielding problem used in MCNP's verification and validation suite. It was observed that both codes became slower on the MIC than on a 6-core X5650 CPU, by a factor of 4 for the MPI code and, abnormally, 20 for the OpenMP code, and both exhibited limited capability of strong scaling. (2) We have recently added a Constructive Solid Geometry (CSG) module to our ARCHER code to provide better support for geometry modelling in radiation shielding simulation. The functions of this module are frequently called in the particle random walk process. To identify the performance bottleneck we developed a CSG proxy application and profiled the code using the geometry data from FS7ONNi. The profiling data showed that the code was primarily memory latency bound on the MIC. This study suggests that despite low initial porting e_ort, Monte Carlo codes do not naturally lend themselves to the MIC platform — just like to the GPUs, and that the memory latency problem needs to be addressed in order to achieve decent performance gain.
Introduction to the Latest Version of the Test-Particle Monte Carlo Code Molflow+
Ady, M
2014-01-01
The Test-Particle Monte Carlo code Molflow+ is getting more and more attention from the scientific community needing detailed 3D calculations of vacuum in the molecular flow regime mainly, but not limited to, the particle accelerator field. Substantial changes, bug fixes, geometry-editing and modelling features, and computational speed improvements have been made to the code in the last couple of years. This paper will outline some of these new features, and show examples of applications to the design and analysis of vacuum systems at CERN and elsewhere.
New capabilities for Monte Carlo simulation of deuteron transport and secondary products generation
International Nuclear Information System (INIS)
Sauvan, P.; Sanz, J.; Ogando, F.
2010-01-01
Several important research programs are dedicated to the development of facilities based on deuteron accelerators. In designing these facilities, the definition of a validated computational approach able to simulate deuteron transport and evaluate deuteron interactions and production of secondary particles with acceptable precision is a very important issue. Current Monte Carlo codes, such as MCNPX or PHITS, when applied for deuteron transport calculations use built-in semi-analytical models to describe deuteron interactions. These models are found unreliable in predicting neutron and photon generated by low energy deuterons, typically present in those facilities. We present a new computational tool, resulting from an extension of the MCNPX code, which improve significantly the treatment of problems where any secondary product (neutrons, photons, tritons, etc.) generated by low energy deuterons reactions could play a major role. Firstly, it handles deuteron evaluated data libraries, which allow describing better low deuteron energy interactions. Secondly, it includes a reduction variance technique for production of secondary particles by charged particle-induced nuclear interactions, which allow reducing drastically the computing time needed in transport and nuclear response calculations. Verification of the computational tool is successfully achieved. This tool can be very helpful in addressing design issues such as selection of the dedicated neutron production target and accelerator radioprotection analysis. It can be also helpful to test the deuteron cross-sections under development in the frame of different international nuclear data programs.
Positron follow-up in liquid water: I. A new Monte Carlo track-structure code.
Champion, C; Le Loirec, C
2006-04-07
When biological matter is irradiated by charged particles, a wide variety of interactions occur, which lead to a deep modification of the cellular environment. To understand the fine structure of the microscopic distribution of energy deposits, Monte Carlo event-by-event simulations are particularly suitable. However, the development of these track-structure codes needs accurate interaction cross sections for all the electronic processes: ionization, excitation, positronium formation and even elastic scattering. Under these conditions, we have recently developed a Monte Carlo code for positrons in water, the latter being commonly used to simulate the biological medium. All the processes are studied in detail via theoretical differential and total cross-section calculations performed by using partial wave methods. Comparisons with existing theoretical and experimental data in terms of stopping powers, mean energy transfers and ranges show very good agreements. Moreover, thanks to the theoretical description of positronium formation, we have access, for the first time, to the complete kinematics of the electron capture process. Then, the present Monte Carlo code is able to describe the detailed positronium history, which will provide useful information for medical imaging (like positron emission tomography) where improvements are needed to define with the best accuracy the tumoural volumes.
Energy Technology Data Exchange (ETDEWEB)
Bellezzo, Murillo
2014-09-01
As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo Method (MCM) has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this thesis, the CUBMC code is presented, a GPU-based MC photon transport algorithm for dose calculation under the Compute Unified Device Architecture (CUDA) platform. The simulation of physical events is based on the algorithm used in PENELOPE, and the cross section table used is the one generated by the MATERIAL routine, also present in PENELOPE code. Photons are transported in voxel-based geometries with different compositions. There are two distinct approaches used for transport simulation. The rst of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon ignores the existence of borders and travels in homogeneous fictitious media. The CUBMC code aims to be an alternative of Monte Carlo simulator code that, by using the capability of parallel processing of graphics processing units (GPU), provide high performance simulations in low cost compact machines, and thus can be applied in clinical cases and incorporated in treatment planning systems for radiotherapy. (author)
Energy Technology Data Exchange (ETDEWEB)
Sjenitzer, Bart L.; Hoogenboom, J. Eduard, E-mail: B.L.Sjenitzer@TUDelft.nl, E-mail: J.E.Hoogenboom@TUDelft.nl [Delft University of Technology (Netherlands)
2011-07-01
A new Dynamic Monte Carlo method is implemented in the general purpose Monte Carlo code Tripoli 4.6.1. With this new method incorporated, a general purpose code can be used for safety transient analysis, such as the movement of a control rod or in an accident scenario. To make the Tripoli code ready for calculating on dynamic systems, the Tripoli scheme had to be altered to incorporate time steps, to include the simulation of delayed neutron precursors and to simulate prompt neutron chains. The modified Tripoli code is tested on two sample cases, a steady-state system and a subcritical system and the resulting neutron fluxes behave just as expected. The steady-state calculation has a constant neutron flux over time and this result shows the stability of the calculation. The neutron flux stays constant with acceptable variance. This also shows that the starting conditions are determined correctly. The sub-critical case shows that the code can also handle dynamic systems with a varying neutron flux. (author)
International Nuclear Information System (INIS)
Sjenitzer, Bart L.; Hoogenboom, J. Eduard
2011-01-01
A new Dynamic Monte Carlo method is implemented in the general purpose Monte Carlo code Tripoli 4.6.1. With this new method incorporated, a general purpose code can be used for safety transient analysis, such as the movement of a control rod or in an accident scenario. To make the Tripoli code ready for calculating on dynamic systems, the Tripoli scheme had to be altered to incorporate time steps, to include the simulation of delayed neutron precursors and to simulate prompt neutron chains. The modified Tripoli code is tested on two sample cases, a steady-state system and a subcritical system and the resulting neutron fluxes behave just as expected. The steady-state calculation has a constant neutron flux over time and this result shows the stability of the calculation. The neutron flux stays constant with acceptable variance. This also shows that the starting conditions are determined correctly. The sub-critical case shows that the code can also handle dynamic systems with a varying neutron flux. (author)
Application of SN and Monte Carlo codes to the SHEBA critical assemblies
International Nuclear Information System (INIS)
O'Dell, R.D.
1993-01-01
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 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
Multiple component codes based generalized LDPC codes for high-speed optical transport.
Djordjevic, Ivan B; Wang, Ting
2014-07-14
A class of generalized low-density parity-check (GLDPC) codes suitable for optical communications is proposed, which consists of multiple local codes. It is shown that Hamming, BCH, and Reed-Muller codes can be used as local codes, and that the maximum a posteriori probability (MAP) decoding of these local codes by Ashikhmin-Lytsin algorithm is feasible in terms of complexity and performance. We demonstrate that record coding gains can be obtained from properly designed GLDPC codes, derived from multiple component codes. We then show that several recently proposed classes of LDPC codes such as convolutional and spatially-coupled codes can be described using the concept of GLDPC coding, which indicates that the GLDPC coding can be used as a unified platform for advanced FEC enabling ultra-high speed optical transport. The proposed class of GLDPC codes is also suitable for code-rate adaption, to adjust the error correction strength depending on the optical channel conditions.
International Nuclear Information System (INIS)
Kim, Jung-Ha; Hill, Robin; Kuncic, Zdenka
2012-01-01
The Monte Carlo (MC) method has proven invaluable for radiation transport simulations to accurately determine radiation doses and is widely considered a reliable computational measure that can substitute a physical experiment where direct measurements are not possible or feasible. In the EGSnrc/BEAMnrc MC codes, there are several user-specified parameters and customized transport algorithms, which may affect the calculation results. In order to fully utilize the MC methods available in these codes, it is essential to understand all these options and to use them appropriately. In this study, the effects of the electron transport algorithms in EGSnrc/BEAMnrc, which are often a trade-off between calculation accuracy and efficiency, were investigated in the buildup region of a homogeneous water phantom and also in a heterogeneous phantom using the DOSRZnrc user code. The algorithms and parameters investigated include: boundary crossing algorithm (BCA), skin depth, electron step algorithm (ESA), global electron cutoff energy (ECUT) and electron production cutoff energy (AE). The variations in calculated buildup doses were found to be larger than 10% for different user-specified transport parameters. We found that using BCA = EXACT gave the best results in terms of accuracy and efficiency in calculating buildup doses using DOSRZnrc. In addition, using the ESA = PRESTA-I option was found to be the best way of reducing the total calculation time without losing accuracy in the results at high energies (few keV ∼ MeV). We also found that although choosing a higher ECUT/AE value in the beam modelling can dramatically improve computation efficiency, there is a significant trade-off in surface dose uncertainty. Our study demonstrates that a careful choice of user-specified transport parameters is required when conducting similar MC calculations. (note)
International Nuclear Information System (INIS)
Zitouni, Y.
1987-04-01
In the field of shielding, the requirement of radiation transport calculations in severe conditions, characterized by irreducible three-dimensional geometries has increased the use of the Monte Carlo method. The latter has proved to be the only rigorous and appropriate calculational method in such conditions. However, further efforts at optimization are still necessary to render the technique practically efficient, despite recent improvements in the Monte Carlo codes, the progress made in the field of computers and the availability of accurate nuclear data. Moreover, the personal experience acquired in the field and the control of sophisticated calculation procedures are of the utmost importance. The aim of the work which has been carried out is the gathering of all the necessary elements and features that would lead to an efficient utilization of the Monte Carlo method used in connection with shielding problems. The study of the general aspects of the method and the exploitation techniques of the MORSE code, which has proved to be one of the most comprehensive of the Monte Carlo codes, lead to a successful analysis of an actual case. In fact, the severe conditions and difficulties met have been overcome using such a stochastic simulation code. Finally, a critical comparison between calculated and high-accuracy experimental results has allowed the final confirmation of the methodology used by us
Dewaraja, Yuni K; Ljungberg, Michael; Majumdar, Amitava; Bose, Abhijit; Koral, Kenneth F
2002-02-01
This paper reports the implementation of the SIMIND Monte Carlo code on an IBM SP2 distributed memory parallel computer. Basic aspects of running Monte Carlo particle transport calculations on parallel architectures are described. Our parallelization is based on equally partitioning photons among the processors and uses the Message Passing Interface (MPI) library for interprocessor communication and the Scalable Parallel Random Number Generator (SPRNG) to generate uncorrelated random number streams. These parallelization techniques are also applicable to other distributed memory architectures. A linear increase in computing speed with the number of processors is demonstrated for up to 32 processors. This speed-up is especially significant in Single Photon Emission Computed Tomography (SPECT) simulations involving higher energy photon emitters, where explicit modeling of the phantom and collimator is required. For (131)I, the accuracy of the parallel code is demonstrated by comparing simulated and experimental SPECT images from a heart/thorax phantom. Clinically realistic SPECT simulations using the voxel-man phantom are carried out to assess scatter and attenuation correction.
Energy Technology Data Exchange (ETDEWEB)
Parreno Z, F.; Paucar J, R.; Picon C, C. [Instituto Peruano de Energia Nuclear, Av. Canada 1470, San Borja, Lima 41 (Peru)
1998-12-31
The simulation by Monte Carlo is tool which Medical Physics counts with it for the development of its research, the interest by this tool is growing, as we may observe in the main scientific journals for the years 1995-1997 where more than 27 % of the papers treat over Monte Carlo and/or its applications in the radiation transport.In the Peruvian Institute of Nuclear Energy we are implementing and making use of the MCNP4 and EGS4 codes. In this work are presented the general features of the Monte Carlo method and its more useful applications in Medical Physics. Likewise, it is made a simulation of the calculation of isodose curves in an interstitial treatment with Ir-192 wires in a mammary gland carcinoma. (Author)
Temperature variance study in Monte-Carlo photon transport theory
International Nuclear Information System (INIS)
Giorla, J.
1985-10-01
We study different Monte-Carlo methods for solving radiative transfer problems, and particularly Fleck's Monte-Carlo method. We first give the different time-discretization schemes and the corresponding stability criteria. Then we write the temperature variance as a function of the variances of temperature and absorbed energy at the previous time step. Finally we obtain some stability criteria for the Monte-Carlo method in the stationary case [fr
Performance of the improved version of Monte Carlo code A 3MCNP for large-scale shielding problems
International Nuclear Information System (INIS)
Omura, M.; Miyake, Y.; Hasegawa, T.; Ueki, K.; Sato, O.; Haghighat, A.; Sjoden, G. E.
2005-01-01
A 3MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, which automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic 'importance' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3MCNP uses the three-dimensional (3-D) Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A 3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3MCNP (referred to as A 3MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3MCNPV for a concrete cask neutron and gamma-ray shielding problem, and a PWR dosimetry problem. (authors)
Monte Carlo simulation of a coded-aperture thermal neutron camera
International Nuclear Information System (INIS)
Dioszegi, I.; Salwen, C.; Forman, L.
2011-01-01
We employed the MCNPX Monte Carlo code to simulate image formation in a coded-aperture thermal-neutron camera. The camera, developed at Brookhaven National Laboratory (BNL), consists of a 20 x 17 cm"2 active area "3He-filled position-sensitive wire chamber in a cadmium enclosure box. The front of the box is a coded-aperture cadmium mask (at present with three different resolutions). We tested the detector experimentally with various arrangements of moderated point-neutron sources. The purpose of using the Monte Carlo modeling was to develop an easily modifiable model of the device to predict the detector's behavior using different mask patterns, and also to generate images of extended-area sources or large numbers (up to ten) of them, that is important for nonproliferation and arms-control verification, but difficult to achieve experimentally. In the model, we utilized the advanced geometry capabilities of the MCNPX code to simulate the coded aperture mask. Furthermore, the code simulated the production of thermal neutrons from fission sources surrounded by a thermalizer. With this code we also determined the thermal-neutron shadow cast by the cadmium mask; the calculations encompassed fast- and epithermal-neutrons penetrating into the detector through the mask. Since the process of signal production in "3He-filled position-sensitive wire chambers is well known, we omitted this part from our modeling. Simplified efficiency values were used for the three (thermal, epithermal, and fast) neutron-energy regions. Electronic noise and the room's background were included as a uniform irradiation component. We processed the experimental- and simulated-images using identical LabVIEW virtual instruments. (author)
Energy Technology Data Exchange (ETDEWEB)
Mendenhall, Marcus H., E-mail: marcus.h.mendenhall@vanderbilt.edu [Vanderbilt University, Department of Electrical Engineering, P.O. Box 351824B, Nashville, TN 37235 (United States); Weller, Robert A., E-mail: robert.a.weller@vanderbilt.edu [Vanderbilt University, Department of Electrical Engineering, P.O. Box 351824B, Nashville, TN 37235 (United States)
2012-03-01
In Monte Carlo particle transport codes, it is often important to adjust reaction cross-sections to reduce the variance of calculations of relatively rare events, in a technique known as non-analog Monte Carlo. We present the theory and sample code for a Geant4 process which allows the cross-section of a G4VDiscreteProcess to be scaled, while adjusting track weights so as to mitigate the effects of altered primary beam depletion induced by the cross-section change. This makes it possible to increase the cross-section of nuclear reactions by factors exceeding 10{sup 4} (in appropriate cases), without distorting the results of energy deposition calculations or coincidence rates. The procedure is also valid for bias factors less than unity, which is useful in problems that involve the computation of particle penetration deep into a target (e.g. atmospheric showers or shielding studies).
International Nuclear Information System (INIS)
Mendenhall, Marcus H.; Weller, Robert A.
2012-01-01
In Monte Carlo particle transport codes, it is often important to adjust reaction cross-sections to reduce the variance of calculations of relatively rare events, in a technique known as non-analog Monte Carlo. We present the theory and sample code for a Geant4 process which allows the cross-section of a G4VDiscreteProcess to be scaled, while adjusting track weights so as to mitigate the effects of altered primary beam depletion induced by the cross-section change. This makes it possible to increase the cross-section of nuclear reactions by factors exceeding 10 4 (in appropriate cases), without distorting the results of energy deposition calculations or coincidence rates. The procedure is also valid for bias factors less than unity, which is useful in problems that involve the computation of particle penetration deep into a target (e.g. atmospheric showers or shielding studies).
Improved cache performance in Monte Carlo transport calculations using energy banding
Siegel, A.; Smith, K.; Felker, K.; Romano, P.; Forget, B.; Beckman, P.
2014-04-01
We present an energy banding algorithm for Monte Carlo (MC) neutral particle transport simulations which depend on large cross section lookup tables. In MC codes, read-only cross section data tables are accessed frequently, exhibit poor locality, and are typically too much large to fit in fast memory. Thus, performance is often limited by long latencies to RAM, or by off-node communication latencies when the data footprint is very large and must be decomposed on a distributed memory machine. The proposed energy banding algorithm allows maximal temporal reuse of data in band sizes that can flexibly accommodate different architectural features. The energy banding algorithm is general and has a number of benefits compared to the traditional approach. In the present analysis we explore its potential to achieve improvements in time-to-solution on modern cache-based architectures.
Monte Carlo electron-transport calculations for clinical beams using energy grouping
Energy Technology Data Exchange (ETDEWEB)
Teng, S P; Anderson, D W; Lindstrom, D G
1986-01-01
A Monte Carlo program has been utilized to study the penetration of broad electron beams into a water phantom. The MORSE-E code, originally developed for neutron and photon transport, was chosen for adaptation to electrons because of its versatility. The electron energy degradation model employed logarithmic spacing of electron energy groups and included effects of elastic scattering, inelastic-moderate-energy-loss-processes and inelastic-large-energy-loss-processes (catastrophic). Energy straggling and angular deflections were modeled from group to group, using the Moeller cross section for energy loss, and Goudsmit-Saunderson theory to describe angular deflections. The resulting energy- and electron-deposition distributions in depth were obtained at 10 and 20 MeV and are compared with ETRAN results and broad beam experimental data from clinical accelerators.
SHIELD-HIT12A - a Monte Carlo particle transport program for ion therapy research
DEFF Research Database (Denmark)
Bassler, Niels; Hansen, David Christoffer; Lühr, Armin
2014-01-01
. We experienced that new users quickly learn to use SHIELD-HIT12A and setup new geometries. Contrary to previous versions of SHIELD-HIT, the 12A distribution comes along with easy-to-use example files and an English manual. A new implementation of Vavilov straggling resulted in a massive reduction......Abstract. Purpose: The Monte Carlo (MC) code SHIELD-HIT simulates the transport of ions through matter. Since SHIELD-HIT08 we added numerous features that improves speed, usability and underlying physics and thereby the user experience. The “-A” fork of SHIELD-HIT also aims to attach SHIELD....... It supports native formats compatible with the heavy ion treatment planning system TRiP. Stopping power files follow ICRU standard and are generated using the libdEdx library, which allows the user to choose from a multitude of stopping power tables. Results: SHIELD-HIT12A runs on Linux and Windows platforms...
International Nuclear Information System (INIS)
Sempau, Josep; Wilderman, Scott J.; Bielajew, Alex F.
2000-01-01
A new Monte Carlo (MC) algorithm, the 'dose planning method' (DPM), and its associated computer program for simulating the transport of electrons and photons in radiotherapy class problems employing primary electron beams, is presented. DPM is intended to be a high-accuracy MC alternative to the current generation of treatment planning codes which rely on analytical algorithms based on an approximate solution of the photon/electron Boltzmann transport equation. For primary electron beams, DPM is capable of computing 3D dose distributions (in 1 mm 3 voxels) which agree to within 1% in dose maximum with widely used and exhaustively benchmarked general-purpose public-domain MC codes in only a fraction of the CPU time. A representative problem, the simulation of 1 million 10 MeV electrons impinging upon a water phantom of 128 3 voxels of 1 mm on a side, can be performed by DPM in roughly 3 min on a modern desktop workstation. DPM achieves this performance by employing transport mechanics and electron multiple scattering distribution functions which have been derived to permit long transport steps (of the order of 5 mm) which can cross heterogeneity boundaries. The underlying algorithm is a 'mixed' class simulation scheme, with differential cross sections for hard inelastic collisions and bremsstrahlung events described in an approximate manner to simplify their sampling. The continuous energy loss approximation is employed for energy losses below some predefined thresholds, and photon transport (including Compton, photoelectric absorption and pair production) is simulated in an analogue manner. The δ-scattering method (Woodcock tracking) is adopted to minimize the computational costs of transporting photons across voxels. (author)
International Nuclear Information System (INIS)
Badal, Andreu; Badano, Aldo
2009-01-01
Purpose: It is a known fact that Monte Carlo simulations of radiation transport are computationally intensive and may require long computing times. The authors introduce a new paradigm for the acceleration of Monte Carlo simulations: The use of a graphics processing unit (GPU) as the main computing device instead of a central processing unit (CPU). Methods: A GPU-based Monte Carlo code that simulates photon transport in a voxelized geometry with the accurate physics models from PENELOPE has been developed using the CUDA programming model (NVIDIA Corporation, Santa Clara, CA). Results: An outline of the new code and a sample x-ray imaging simulation with an anthropomorphic phantom are presented. A remarkable 27-fold speed up factor was obtained using a GPU compared to a single core CPU. Conclusions: The reported results show that GPUs are currently a good alternative to CPUs for the simulation of radiation transport. Since the performance of GPUs is currently increasing at a faster pace than that of CPUs, the advantages of GPU-based software are likely to be more pronounced in the future.
Energy Technology Data Exchange (ETDEWEB)
Badal, Andreu; Badano, Aldo [Division of Imaging and Applied Mathematics, OSEL, CDRH, U.S. Food and Drug Administration, Silver Spring, Maryland 20993-0002 (United States)
2009-11-15
Purpose: It is a known fact that Monte Carlo simulations of radiation transport are computationally intensive and may require long computing times. The authors introduce a new paradigm for the acceleration of Monte Carlo simulations: The use of a graphics processing unit (GPU) as the main computing device instead of a central processing unit (CPU). Methods: A GPU-based Monte Carlo code that simulates photon transport in a voxelized geometry with the accurate physics models from PENELOPE has been developed using the CUDA programming model (NVIDIA Corporation, Santa Clara, CA). Results: An outline of the new code and a sample x-ray imaging simulation with an anthropomorphic phantom are presented. A remarkable 27-fold speed up factor was obtained using a GPU compared to a single core CPU. Conclusions: The reported results show that GPUs are currently a good alternative to CPUs for the simulation of radiation transport. Since the performance of GPUs is currently increasing at a faster pace than that of CPUs, the advantages of GPU-based software are likely to be more pronounced in the future.
Badal, Andreu; Badano, Aldo
2009-11-01
It is a known fact that Monte Carlo simulations of radiation transport are computationally intensive and may require long computing times. The authors introduce a new paradigm for the acceleration of Monte Carlo simulations: The use of a graphics processing unit (GPU) as the main computing device instead of a central processing unit (CPU). A GPU-based Monte Carlo code that simulates photon transport in a voxelized geometry with the accurate physics models from PENELOPE has been developed using the CUDATM programming model (NVIDIA Corporation, Santa Clara, CA). An outline of the new code and a sample x-ray imaging simulation with an anthropomorphic phantom are presented. A remarkable 27-fold speed up factor was obtained using a GPU compared to a single core CPU. The reported results show that GPUs are currently a good alternative to CPUs for the simulation of radiation transport. Since the performance of GPUs is currently increasing at a faster pace than that of CPUs, the advantages of GPU-based software are likely to be more pronounced in the future.
Current status of high energy nucleon-meson transport code
Energy Technology Data Exchange (ETDEWEB)
Takada, Hiroshi; Sasa, Toshinobu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
Current status of design code of accelerator (NMTC/JAERI code), outline of physical model and evaluation of accuracy of code were reported. To evaluate the nuclear performance of accelerator and strong spallation neutron origin, the nuclear reaction between high energy proton and target nuclide and behaviors of various produced particles are necessary. The nuclear design of spallation neutron system used a calculation code system connected the high energy nucleon{center_dot}meson transport code and the neutron{center_dot}photon transport code. NMTC/JAERI is described by the particle evaporation process under consideration of competition reaction of intranuclear cascade and fission process. Particle transport calculation was carried out for proton, neutron, {pi}- and {mu}-meson. To verify and improve accuracy of high energy nucleon-meson transport code, data of spallation and spallation neutron fragment by the integral experiment were collected. (S.Y.)
(U) Introduction to Monte Carlo Methods
Energy Technology Data Exchange (ETDEWEB)
Hungerford, Aimee L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2017-03-20
Monte Carlo methods are very valuable for representing solutions to particle transport problems. Here we describe a “cook book” approach to handling the terms in a transport equation using Monte Carlo methods. Focus is on the mechanics of a numerical Monte Carlo code, rather than the mathematical foundations of the method.
Accuracy assessment of a new Monte Carlo based burnup computer code
International Nuclear Information System (INIS)
El Bakkari, B.; ElBardouni, T.; Nacir, B.; ElYounoussi, C.; Boulaich, Y.; Meroun, O.; Zoubair, M.; Chakir, E.
2012-01-01
Highlights: ► A new burnup code called BUCAL1 was developed. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► Validation of BUCAL1 was done by code to code comparison using VVER-1000 LEU Benchmark Assembly. ► Differences from BM value were found to be ± 600 pcm for k ∞ and ±6% for the isotopic compositions. ► The effect on reactivity due to the burnup of Gd isotopes is well reproduced by BUCAL1. - Abstract: This study aims to test for the suitability and accuracy of a new home-made Monte Carlo burnup code, called BUCAL1, by investigating and predicting the neutronic behavior of a “VVER-1000 LEU Assembly Computational Benchmark”, at lattice level. BUCAL1 uses MCNP tally information directly in the computation; this approach allows performing straightforward and accurate calculation without having to use the calculated group fluxes to perform transmutation analysis in a separate code. ENDF/B-VII evaluated nuclear data library was used in these calculations. Processing of the data library is performed using recent updates of NJOY99 system. Code to code comparisons with the reported Nuclear OECD/NEA results are presented and analyzed.
Comparison of TITAN hybrid deterministic transport code and MCNP5 for simulation of SPECT
International Nuclear Information System (INIS)
Royston, K.; Haghighat, A.; Yi, C.
2010-01-01
Traditionally, Single Photon Emission Computed Tomography (SPECT) simulations use Monte Carlo methods. The hybrid deterministic transport code TITAN has recently been applied to the simulation of a SPECT myocardial perfusion study. The TITAN SPECT simulation uses the discrete ordinates formulation in the phantom region and a simplified ray-tracing formulation outside of the phantom. A SPECT model has been created in the Monte Carlo Neutral particle (MCNP)5 Monte Carlo code for comparison. In MCNP5 the collimator is directly modeled, but TITAN instead simulates the effect of collimator blur using a circular ordinate splitting technique. Projection images created using the TITAN code are compared to results using MCNP5 for three collimator acceptance angles. Normalized projection images for 2.97 deg, 1.42 deg and 0.98 deg collimator acceptance angles had maximum relative differences of 21.3%, 11.9% and 8.3%, respectively. Visually the images are in good agreement. Profiles through the projection images were plotted to find that the TITAN results followed the shape of the MCNP5 results with some differences in magnitude. A timing comparison on 16 processors found that the TITAN code completed the calculation 382 to 2787 times faster than MCNP5. Both codes exhibit good parallel performance. (author)
RADTRAN: a computer code to analyze transportation of radioactive material
International Nuclear Information System (INIS)
Taylor, J.M.; Daniel, S.L.
1977-04-01
A computer code is presented which predicts the environmental impact of any specific scheme of radioactive material transportation. Results are presented in terms of annual latent cancer fatalities and annual early fatility probability resulting from exposure, during normal transportation or transport accidents. The code is developed in a generalized format to permit wide application including normal transportation analysis; consideration of alternatives; and detailed consideration of specific sectors of industry
R and D on automatic modeling methods for Monte Carlo codes FLUKA
International Nuclear Information System (INIS)
Wang Dianxi; Hu Liqin; Wang Guozhong; Zhao Zijia; Nie Fanzhi; Wu Yican; Long Pengcheng
2013-01-01
FLUKA is a fully integrated particle physics Monte Carlo simulation package. It is necessary to create the geometry models before calculation. However, it is time- consuming and error-prone to describe the geometry models manually. This study developed an automatic modeling method which could automatically convert computer-aided design (CAD) geometry models into FLUKA models. The conversion program was integrated into CAD/image-based automatic modeling program for nuclear and radiation transport simulation (MCAM). Its correctness has been demonstrated. (authors)
International Nuclear Information System (INIS)
Su, L.; Du, X.; Liu, T.; Xu, X. G.
2013-01-01
An electron-photon coupled Monte Carlo code ARCHER - Accelerated Radiation-transport Computations in Heterogeneous EnviRonments - is being developed at Rensselaer Polytechnic Institute as a software test-bed for emerging heterogeneous high performance computers that utilize accelerators such as GPUs (Graphics Processing Units). This paper presents the preliminary code development and the testing involving radiation dose related problems. In particular, the paper discusses the electron transport simulations using the class-II condensed history method. The considered electron energy ranges from a few hundreds of keV to 30 MeV. As for photon part, photoelectric effect, Compton scattering and pair production were simulated. Voxelized geometry was supported. A serial CPU (Central Processing Unit)code was first written in C++. The code was then transplanted to the GPU using the CUDA C 5.0 standards. The hardware involved a desktop PC with an Intel Xeon X5660 CPU and six NVIDIA Tesla M2090 GPUs. The code was tested for a case of 20 MeV electron beam incident perpendicularly on a water-aluminum-water phantom. The depth and later dose profiles were found to agree with results obtained from well tested MC codes. Using six GPU cards, 6*10 6 electron histories were simulated within 2 seconds. In comparison, the same case running the EGSnrc and MCNPX codes required 1645 seconds and 9213 seconds, respectively. On-going work continues to test the code for different medical applications such as radiotherapy and brachytherapy. (authors)
Sensitivity analysis of the titan hybrid deterministic transport code for SPECT simulation
International Nuclear Information System (INIS)
Royston, Katherine K.; Haghighat, Alireza
2011-01-01
Single photon emission computed tomography (SPECT) has been traditionally simulated using Monte Carlo methods. The TITAN code is a hybrid deterministic transport code that has recently been applied to the simulation of a SPECT myocardial perfusion study. For modeling SPECT, the TITAN code uses a discrete ordinates method in the phantom region and a combined simplified ray-tracing algorithm with a fictitious angular quadrature technique to simulate the collimator and generate projection images. In this paper, we compare the results of an experiment with a physical phantom with predictions from the MCNP5 and TITAN codes. While the results of the two codes are in good agreement, they differ from the experimental data by ∼ 21%. In order to understand these large differences, we conduct a sensitivity study by examining the effect of different parameters including heart size, collimator position, collimator simulation parameter, and number of energy groups. (author)
International Nuclear Information System (INIS)
Li, Junli; Qiu, Rui; Yan, Congchong; Xie, Wenzhang; Zeng, Zhi; Li, Chunyan; Wu, Zhen; Tung, Chuanjong
2015-01-01
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 (authors)
Variation in computer time with geometry prescription in monte carlo code KENO-IV
International Nuclear Information System (INIS)
Gopalakrishnan, C.R.
1988-01-01
In most studies, the Monte Carlo criticality code KENO-IV has been compared with other Monte Carlo codes, but evaluation of its performance with different box descriptions has not been done so far. In Monte Carlo computations, any fractional savings of computing time is highly desirable. Variation in computation time with box description in KENO for two different fast reactor fuel subassemblies of FBTR and PFBR is studied. The K eff of an infinite array of fuel subassemblies is calculated by modelling the subassemblies in two different ways (i) multi-region, (ii) multi-box. In addition to these two cases, excess reactivity calculations of FBTR are also performed in two ways to study this effect in a complex geometry. It is observed that the K eff values calculated by multi-region and multi-box models agree very well. However the increase in computation time from the multi-box to the multi-region is considerable, while the difference in computer storage requirements for the two models is negligible. This variation in computing time arises from the way the neutron is tracked in the two cases. (author)
The use of Monte-Carlo codes for treatment planning in external-beam radiotherapy
International Nuclear Information System (INIS)
Alan, E.; Nahum, PhD.
2003-01-01
Monte Carlo simulation of radiation transport is a very powerful technique. There are basically no exact solutions to the Boltzmann transport equation. Even, the 'straightforward' situation (in radiotherapy) of an electron beam depth-dose distribution in water proves to be too difficult for analytical methods without making gross approximations such as ignoring energy-loss straggling, large-angle single scattering and Bremsstrahlung production. monte Carlo is essential when radiation is transport from one medium into another. As the particle (be it a neutron, photon, electron, proton) crosses the boundary then a new set of interaction cross-sections is simply read in and the simulation continues as though the new medium were infinite until the next boundary is encountered. Radiotherapy involves directing a beam of megavoltage x rays or electrons (occasionally protons) at a very complex object, the human body. Monte Carlo simulation has proved in valuable at many stages of the process of accurately determining the distribution of absorbed dose in the patient. Some of these applications will be reviewed here. (Rogers and al 1990; Andreo 1991; Mackie 1990). (N.C.)
Energy Technology Data Exchange (ETDEWEB)
Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)
2016-09-15
A burn-up calculation of large systems by Monte-Carlo code (MCU) is complex process and it requires large computational costs. Previously prepared isotopic compositions are proposed to be used for the Monte-Carlo code calculations of different system states with burnt fuel. Isotopic compositions are calculated by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by the engineering codes (TVS-M, BIPR-7A and PERMAK-A). The multiplication factors and power distributions of FAs from a 3-D reactor core are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The separate conditions of the burnt core are observed. The results of MCU calculations were compared with those that were obtained by engineering codes.
Comment on ‘egs_brachy: a versatile and fast Monte Carlo code for brachytherapy’
Yegin, Gultekin
2018-02-01
In a recent paper (Chamberland et al 2016 Phys. Med. Biol. 61 8214) develop a new Monte Carlo code called egs_brachy for brachytherapy treatments. It is based on EGSnrc, and written in the C++ programming language. In order to benchmark the egs_brachy code, the authors use it in various test case scenarios in which complex geometry conditions exist. Another EGSnrc based brachytherapy dose calculation engine, BrachyDose, is used for dose comparisons. The authors fail to prove that egs_brachy can produce reasonable dose values for brachytherapy sources in a given medium. The dose comparisons in the paper are erroneous and misleading. egs_brachy should not be used in any further research studies unless and until all the potential bugs are fixed in the code.
International Nuclear Information System (INIS)
1982-01-01
1 - Description of problem or function: Format: SAIL format; Number of groups: 23 neutron / 17 gamma-ray; Nuclides: Type 04 Concrete and Low Carbon Steel (A533B). Origin: Science Applications, Inc (SAI); Weighting spectrum: yes. SAIL is a library of albedo scattering data to be used in three-dimensional Monte Carlo codes to solve radiation transport problems specific to the reactor pressure vessel cavity region of a LWR. The library contains data for Type 04 Concrete and Low Carbon Steel (A533B). 2 - Method of solution: The calculation of the albedo data was perform- ed with a version of the discrete ordinates transport code DOT which treats the transport of neutrons, secondary gamma-rays and gamma- rays in one dimension, while maintaining the complete two-dimension- al treatment of the angular dependence
International Nuclear Information System (INIS)
Nouri, A.
1994-01-01
Criticality studies in nuclear fuel cycle are based on Monte Carlo method. These codes use multigroup cross sections which can verify by experimental configurations or by use of reference codes such Tripoli 2. In this Tripoli 2 code nuclear data are errors attached and asked for experimental studies with critical experiences. This is one of the aim of this thesis. To calculate the keff of interacted fissile units we have used the multigroup Monte Carlo code Moret with convergence problems. A new estimator of reactions rates permit to better approximate the neutrons exchange between units and a new importance function has been tested. 2 annexes
International Nuclear Information System (INIS)
Roux, A.
2001-01-01
The diversity of radiological accidents makes difficult the medical prognosis and the therapy choice from only clinical observations. To complete this information, it is important to know the global dose received by the organism and the dose distributions in depth in tissues. The dose estimation can be made by a physical reconstruction of the accident with the help of tools based on experimental techniques or on calculation. The software of the geometry construction (M.G.E.D.), associated to the Monte-Carlo code of photons and neutrons transport (M.O.R.S.E.) replies these constraints. An important result of this work is to determine the principal parameters to know in function of the accident type, as well as the precision level required for these parameters. (N.C.)
New scope covered by PHITS. Particle and heavy ion transport code system
International Nuclear Information System (INIS)
Nakamura, Takashi; Niita, Koji; Iwase, Hiroshi; Sato, Tatsuhiko
2006-01-01
PHITS is a general high energy transport calculation code from hadron to heavy ions, which embedded in NMTC-JAM with JQMD code. Outline of PHITS and many application examples are stated. PHITS has been used by the shielding calculations of J-PARC, GSI, RIA and Big-RIPS and the good results were reported. The evaluation of exposure dose of astronauts, airmen, proton and heavy ion therapy, and estimation of error frequency of semiconductor software are explained as the application examples. Relation between the event generator and Monte Carlo method and the future are described. (S.Y.)