Morse Monte Carlo Radiation Transport Code System
Emmett, M.B.
1975-02-01
The report contains sections containing descriptions of the MORSE and PICTURE codes, input descriptions, sample problems, deviations of the physical equations and explanations of the various error messages. The MORSE code is a multipurpose neutron and gamma-ray transport Monte Carlo code. Time dependence for both shielding and criticality problems is provided. General three-dimensional geometry may be used with an albedo option available at any material surface. The PICTURE code provide aid in preparing correct input data for the combinatorial geometry package CG. It provides a printed view of arbitrary two-dimensional slices through the geometry. By inspecting these pictures one may determine if the geometry specified by the input cards is indeed the desired geometry. 23 refs. (WRF)
MORSE Monte Carlo radiation transport code system
This report is an addendum to the MORSE report, ORNL-4972, originally published in 1975. This addendum contains descriptions of several modifications to the MORSE Monte Carlo Code, replacement pages containing corrections, Part II of the report which was previously unpublished, and a new Table of Contents. The modifications include a Klein Nishina estimator for gamma rays. Use of such an estimator required changing the cross section routines to process pair production and Compton scattering cross sections directly from ENDF tapes and writing a new version of subroutine RELCOL. Another modification is the use of free form input for the SAMBO analysis data. This required changing subroutines SCORIN and adding new subroutine RFRE. References are updated, and errors in the original report have been corrected
MORSE Monte Carlo radiation transport code system
For a number of years the MORSE user community has requested additional help in setting up problems using various options. The sample problems distributed with MORSE did not fully demonstrate the capability of the code. At Oak Ridge National Laboratory the code originators had a complete set of sample problems, but funds for documenting and distributing them were never available. Recently the number of requests for listings of input data and results for running some particular option the user was trying to implement has increased to the point where it is not feasible to handle them on an individual basis. Consequently it was decided to package a set of sample problems which illustrates more adequately how to run MORSE. This write-up may be added to Part III of the MORSE report. These sample problems include a combined neutron-gamma case, a neutron only case, a gamma only case, an adjoint case, a fission case, a time-dependent fission case, the collision density case, an XCHEKR run and a PICTUR run
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)
Monte Carlo Code System Development for Liquid Metal Reactor
Kim, Chang Hyo; Shim, Hyung Jin; Han, Beom Seok; Park, Ho Jin; Park, Dong Gyu [Seoul National University, Seoul (Korea, Republic of)
2007-03-15
We have implemented the composition cell class and the use cell to MCCARD for hierarchy input processing. For the inputs of KALlMER-600 core consisted of 336 assemblies, we require the geometric data of 91,056 pin cells. Using hierarchy input processing, it was observed that the system geometries are correctly handled with the geometric data of total 611 cells; 2 cells for fuel rods, 2 cells for guide holes, 271 translation cells for rods, and 336 translation cells for assemblies. We have developed monte carlo decay-chain models based on decay chain model of REBUS code for liquid metal reactor analysis. Using developed decay-chain models, the depletion analysis calculations have performed for the homogeneous and heterogeneous model of KALlMER-600. The k-effective for the depletion analysis agrees well with that of REBUS code. and the developed decay chain models shows more efficient performance for time and memories, as compared with the existing decay chain model The chi-square criterion has been developed to diagnose the temperature convergence for the MC TjH feedback calculations. From the application results to the KALlMER pin and fuel assembly problem, it is observed that the new criterion works well Wc have applied the high efficiency variance reduction technique by splitting Russian roulette to estimate the PPPF of the KALIMER core at BOC. The PPPF of KALlMER core at BOC is 1.235({+-}0.008). The developed technique shows four time faster calculation, as compared with the existin2 calculation Subject Keywords Monte Carlo
Monte Carlo capabilities of the SCALE code system
Highlights: • Foundational Monte Carlo capabilities of SCALE are described. • Improvements in continuous-energy treatments are detailed. • New methods for problem-dependent temperature corrections are described. • New methods for sensitivity analysis and depletion are described. • Nuclear data, users interfaces, and quality assurance activities are summarized. - Abstract: SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a “plug-and-play” framework that includes three deterministic and three Monte Carlo radiation transport solvers that can be selected based on the desired solution, including hybrid deterministic/Monte Carlo simulations. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE’s graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2 will provide several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, and sensitivity and uncertainty analysis. An overview of the Monte Carlo capabilities of SCALE is provided here, with emphasis on new features for SCALE 6.2
Monte Carlo Capabilities of the SCALE Code System
Rearden, B. T.; Petrie, L. M.; Peplow, D. E.; Bekar, K. B.; Wiarda, D.; Celik, C.; Perfetti, C. M.; Ibrahim, A. M.; Hart, S. W. D.; Dunn, M. E.
2014-06-01
SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a "plug-and-play" framework that includes three deterministic and three Monte Carlo radiation transport solvers that can be selected based on the desired solution, including hybrid deterministic/Monte Carlo simulations. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE's graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2, to be released in 2014, will provide several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, and sensitivity and uncertainty analysis. An overview of the Monte Carlo capabilities of SCALE is provided here, with emphasis on new features for SCALE 6.2.
Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)
The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided
Computational Monte Carlo (MC) codes have been used for simulation of nuclear installations mainly for internal monitoring of workers, the well known as Whole Body Counters (WBC). The main goal of this project was the modeling and simulation of the counting efficiency (CE) of a WBC system using three different MC codes: MCNPX, EGSnrc and VMC in-vivo. The simulations were performed for three different groups of analysts. The results shown differences between the three codes, as well as in the results obtained by the same code and modeled by different analysts. Moreover, all the results were also compared to the experimental results obtained in laboratory for meaning of validation and final comparison. In conclusion, it was possible to detect the influence on the results when the system is modeled by different analysts using the same MC code and in which MC code the results were best suited, when comparing to the experimental data result. (author)
Penelope - A code system for Monte Carlo simulation of electron and photon transport
The computer code system PENELOPE (version 2001) performs Monte Carlo simulation of coupled electron-photon transport in arbitrary materials for a wide energy range, from a few hundred eV to about 1 GeV. Photon transport is simulated by means of the standard, detailed simulation scheme. Electron and positron histories are generated on the basis of a mixed procedure, which combines detailed simulation of hard events with condensed simulation of soft interactions. A geometry package called PENGEOM permits the generation of random electron-photon showers in material systems consisting of homogeneous bodies limited by quadric surfaces, i.e. planes, spheres, cylinders, etc. This report is intended not only to serve as a manual of the PENELOPE code system, but also to provide the user with the necessary information to understand the details of the Monte-Carlo algorithm. (authors)
In the design of the incore thermionic reactor system developed under the Advanced Thermionic Initiative (ATI), the fuel is highly enriched uranium dioxide and the moderating medium is zirconium hydride. The traditional burnup and fuel depletion analysis codes have been found to be inadequate for these calculations, largely because of the material and geometry modeled and because the neutron spectra assumed for the codes such as LEOPARD and ORIGEN do not even closely fit that for a small, thermal reactor using ZrH as moderator. More sophisticated codes such as the transport lattice type code WIMS often lack some materials, such as ZrH. Thus a new method which could accurately calculate the neutron spectrum and the appropriate reaction rates within the fuel element is needed. The method developed utilizes and interconnects the accuracy of the Monte Carlo Neutron/Photon (MCNP) method to calculate reaction rates for the important isotopes, and a time dependent depletion routine to calculate the temporal effects on isotope concentrations. This effort required the modification of MCNP itself to perform the additional task of accomplishing burnup calculations. The modified version called, MCNPBURN, evolved to be a general dual purpose code which can be used for standard calculations as well as for burn-up
Implementation of the probability table method in a continuous-energy Monte Carlo code system
Sutton, T.M.; Brown, F.B. [Lockheed Martin Corp., Schenectady, NY (United States)
1998-10-01
RACER is a particle-transport Monte Carlo code that utilizes a continuous-energy treatment for neutrons and neutron cross section data. Until recently, neutron cross sections in the unresolved resonance range (URR) have been treated in RACER using smooth, dilute-average representations. This paper describes how RACER has been modified to use probability tables to treat cross sections in the URR, and the computer codes that have been developed to compute the tables from the unresolved resonance parameters contained in ENDF/B data files. A companion paper presents results of Monte Carlo calculations that demonstrate the effect of the use of probability tables versus the use of dilute-average cross sections for the URR. The next section provides a brief review of the probability table method as implemented in the RACER system. The production of the probability tables for use by RACER takes place in two steps. The first step is the generation of probability tables from the nuclear parameters contained in the ENDF/B data files. This step, and the code written to perform it, are described in Section 3. The tables produced are at energy points determined by the ENDF/B parameters and/or accuracy considerations. The tables actually used in the RACER calculations are obtained in the second step from those produced in the first. These tables are generated at energy points specific to the RACER calculation. Section 4 describes this step and the code written to implement it, as well as modifications made to RACER to enable it to use the tables. Finally, some results and conclusions are presented in Section 5.
Full modelling of the MOSAIC animal PET system based on the GATE Monte Carlo simulation code
Positron emission tomography (PET) systems dedicated to animal imaging are now widely used for biological studies. The scanner performance strongly depends on the design and the characteristics of the system. Many parameters must be optimized like the dimensions and type of crystals, geometry and field-of-view (FOV), sampling, electronics, lightguide, shielding, etc. Monte Carlo modelling is a powerful tool to study the effect of each of these parameters on the basis of realistic simulated data. Performance assessment in terms of spatial resolution, count rates, scatter fraction and sensitivity is an important prerequisite before the model can be used instead of real data for a reliable description of the system response function or for optimization of reconstruction algorithms. The aim of this study is to model the performance of the Philips Mosaic(TM) animal PET system using a comprehensive PET simulation code in order to understand and describe the origin of important factors that influence image quality. We use GATE, a Monte Carlo simulation toolkit for a realistic description of the ring PET model, the detectors, shielding, cap, electronic processing and dead times. We incorporate new features to adjust signal processing to the Anger logic underlying the Mosaic(TM) system. Special attention was paid to dead time and energy spectra descriptions. Sorting of simulated events in a list mode format similar to the system outputs was developed to compare experimental and simulated sensitivity and scatter fractions for different energy thresholds using various models of phantoms describing rat and mouse geometries. Count rates were compared for both cylindrical homogeneous phantoms. Simulated spatial resolution was fitted to experimental data for 18F point sources at different locations within the FOV with an analytical blurring function for electronic processing effects. Simulated and measured sensitivities differed by less than 3%, while scatter fractions agreed
Improvements in the Monte Carlo code for simulating 4πβ(PC)–γ coincidence system measurements
Dias, M.S., E-mail: msdias@ipen.br [Instituto de Pesquisas Energéticas e Nucleares, IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242, 05508-000 São Paulo, SP (Brazil); Takeda, M.N. [Universidade Santo Amaro, UNISA Rua Prof. Enéas da Siqueira Neto 340, 04829-300 São Paulo, SP (Brazil); Toledo, F.; Brancaccio, F.; Tongu, M.L.O.; Koskinas, M.F. [Instituto de Pesquisas Energéticas e Nucleares, IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242, 05508-000 São Paulo, SP (Brazil)
2013-01-11
A Monte Carlo simulation code known as ESQUEMA has been developed by the Nuclear Metrology Laboratory (Laboratório de Metrologia Nuclear—LMN) in the Nuclear and Energy Research Institute (Instituto de Pesquisas Energéticas e Nucleares—IPEN) to be used as a benchmark for radionuclide standardization. The early version of this code simulated only β−γ and ec−γ emitters with reasonably high electron and X-ray energies. To extend the code to include other radionuclides and enable the code to be applied to software coincidence counting systems, several improvements have been made and are presented in this work. -- Highlights: ► Improvements to the Monte Carlo code ESQUEMA are described. ► The experimental extrapolation curve was compared to Monte Carlo simulation. ► Eu-152 was standardized by 4π(PC)β-γ coincidence system and compared to Monte Carlo simulation. ► 4π proportional counter gamma-ray efficiency was calculated by MCNPX and compared with experiment. ► X-ray and positron decay emitters were included in the simulation.
A user`s manual for MASH 1.0: A Monte Carlo Adjoint Shielding Code System
Johnson, J.O. [ed.
1992-03-01
The Monte Carlo Adjoint Shielding Code System, MASH, calculates neutron and gamma-ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air-over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system include the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. MASH is the successor to the Vehicle Code System (VCS) initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the ``dose importance`` of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response a a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user`s manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem (input data and selected output edits) for each code.
A user's manual for MASH 1. 0: A Monte Carlo Adjoint Shielding Code System
Johnson, J.O. (ed.)
1992-03-01
The Monte Carlo Adjoint Shielding Code System, MASH, calculates neutron and gamma-ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air-over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system include the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. MASH is the successor to the Vehicle Code System (VCS) initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the dose importance'' of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response a a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem (input data and selected output edits) for each code.
COG10, Multiparticle Monte Carlo Code System for Shielding and Criticality Use
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
A User's Manual for MASH V1.5 - A Monte Carlo Adjoint Shielding Code System
C. O. Slater; J. M. Barnes; J. O. Johnson; J.D. Drischler
1998-10-01
The Monte Carlo ~djoint ~ielding Code System, MASH, calculates neutron and gamma- ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air- over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system includes the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. The current version, MASH v 1.5, is the successor to the original MASH v 1.0 code system initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the "dose importance" of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response as a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem.
Installation of Monte Carlo neutron and photon transport code system MCNP4
The continuous energy Monte Carlo code MCNP-4 including its graphic functions has been installed on the Sun-4 sparc-2 work station with minor corrections. In order to validate the installed MCNP-4 code, 25 sample problems have been executed on the work station and these results have been compared with the original ones. And, the most of the graphic functions have been demonstrated by using 3 sample problems. Further, additional 14 nuclides have been included to the continuous cross section library edited from JENDL-3. (author)
A 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
SPQR: a Monte Carlo reactor kinetics code
The SPQR Monte Carlo code has been developed to analyze fast reactor core accident problems where conventional methods are considered inadequate. The code is based on the adiabatic approximation of the quasi-static method. This initial version contains no automatic material motion or feedback. An existing Monte Carlo code is used to calculate the shape functions and the integral quantities needed in the kinetics module. Several sample problems have been devised and analyzed. Due to the large statistical uncertainty associated with the calculation of reactivity in accident simulations, the results, especially at later times, differ greatly from deterministic methods. It was also found that in large uncoupled systems, the Monte Carlo method has difficulty in handling asymmetric perturbations
Blazy-Aubignac, L
2007-09-15
The treatment planning systems (T.P.S.) occupy a key position in the radiotherapy service: they realize the projected calculation of the dose distribution and the treatment duration. Traditionally, the quality control of the calculated distribution doses relies on their comparisons with dose distributions measured under the device of treatment. This thesis proposes to substitute these dosimetry measures to the profile of reference dosimetry calculations got by the Penelope Monte-Carlo code. The Monte-Carlo simulations give a broad choice of test configurations and allow to envisage a quality control of dosimetry aspects of T.P.S. without monopolizing the treatment devices. This quality control, based on the Monte-Carlo simulations has been tested on a clinical T.P.S. and has allowed to simplify the quality procedures of the T.P.S.. This quality control, in depth, more precise and simpler to implement could be generalized to every center of radiotherapy. (N.C.)
Highlights: ► New coupled Monte Carlo code system for reference results at operating conditions. ► Automated methodology to create and use temperature-dependent cross section libraries. ► Multi-level coupling scheme between MCNP5 and COBRA-TF with different options. ► Acceleration strategy for coupled Monte Carlo calculations including hybrid approach. ► Sensitivity studies on thermal-scattering models and different sub-channel approaches. -- Abstract: High accuracy code systems are necessary to model core environments with considerable geometry complexity and great material heterogeneity. These features are typical of current and innovative nuclear reactor core designs. Advanced methodologies and state-of-the art coupled code systems must be put into practice in order to model with high accuracy these challenging core designs. The presented research comprises the development and implementation of the thermal–hydraulic feedback to the Monte Carlo method and of speed-up mechanisms to accelerate the Monte Carlo criticality calculation. Coupled Monte-Carlo calculations can serve as reference solutions for verifying high-fidelity coupled deterministic neutron transport methods with detailed and accurate thermal–hydraulic models. The development and verification of such reference high-fidelity coupled multi-physics scheme is performed at the Pennsylvania State University (PSU) in cooperation with AREVA, AREVA NP GmbH in Erlangen, Germany, on the basis of MCNP5, NEM, NJOY and COBRA-TF (CTF) computer codes. This paper presents the latest studies and ameliorations developed to this coupled hybrid system, which includes a new methodology for generation and interpolation of Temperature-Dependent Thermal Scattering Cross Section Libraries for MCNP5, a comparison between sub-channel approaches, and acceleration schemes.
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 P1-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
The treatment planning systems (T.P.S.) occupy a key position in the radiotherapy service: they realize the projected calculation of the dose distribution and the treatment duration. Traditionally, the quality control of the calculated distribution doses relies on their comparisons with dose distributions measured under the device of treatment. This thesis proposes to substitute these dosimetry measures to the profile of reference dosimetry calculations got by the Penelope Monte-Carlo code. The Monte-Carlo simulations give a broad choice of test configurations and allow to envisage a quality control of dosimetry aspects of T.P.S. without monopolizing the treatment devices. This quality control, based on the Monte-Carlo simulations has been tested on a clinical T.P.S. and has allowed to simplify the quality procedures of the T.P.S.. This quality control, in depth, more precise and simpler to implement could be generalized to every center of radiotherapy. (N.C.)
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.)
High accuracy code systems are necessary to model core environments with considerable geometry complexity and great material heterogeneity. These features are typical of current and innovative nuclear reactor core designs. Advanced methodologies and state-of-the art coupled code systems must be put into practice in order to model with high accuracy these challenging core designs. The presented research comprises the development and implementation of the thermal-hydraulic feedback to the Monte Carlo method and of speed-up mechanisms to accelerate the Monte Carlo criticality calculation. Coupled Monte-Carlo calculations can serve as reference solutions for verifying high-fidelity coupled deterministic neutron transport methods with detailed and accurate thermal-hydraulic models. The development and verification of such reference high-fidelity coupled multi-physics scheme is performed at the Pennsylvania State University (PSU) in cooperation with AREVA, AREVA NP GmbH in Erlangen, Germany, on the basis of MCNP5, NEM, NJOY and COBRA-TF (CTF) computer codes. This paper presents the latest studies and ameliorations developed to this coupled hybrid system, which includes a new methodology for generation and interpolation of Temperature-Dependent Thermal Scattering Cross Section Libraries for MCNP5, a comparison between sub-channel approaches, and acceleration schemes. (authors)
The RTP is a light-water moderated and pool-type TRIGA MARK II reactor with power capacity of 1MWt. It was built in 1979 and attained the first criticality on 28 June 1982. The RTP was designed mainly for neutron activation analysis, small angle neutron scattering, neutron radiography, radioisotope production, education and training purposes. It uses standard TRIGA fuel developed by General Atomic in which the zirconium hydride moderator is homogeneously combined with enriched uranium. It has a cylindrical core with which possibility of locating 127 of fuel elements. Both of the coolant and moderator uses light water system and the reflector is made of high purity graphite. Because of its relatively small power, it uses natural convection for its cooling system. To ensure the integrity of the core, fuel shuffling have been carried out several times. Until now, there were 12 configurations of the core, the most recent change being in July 2006. This paper will describe the RTP core calculation using the Monte Carlo MVP code system. VP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation in order to have an accurate and fast Monte Carlo simulation of neutron and photon transport problems. The MVP Monte Carlo 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. When compared to the conventional scalar method, this code could achieve higher computation speed by up to a factor of 10 on the vector super-computer. The RTP core has been modelled using cylinder geometry along the z-coordinate geometry with the MVP code system while its material cross section data is calculated beforehand. The JENDL3.3 data library was used in the whole calculation. The objectives of the calculation are to calculate the multiplication factor values (keff), fission density and flux distribution from the tally data. The calculation also
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)
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)
Geometry system used in the General Monte Carlo transport code SPARTAN
The geometry routines used in the general-purpose, three-dimensional particle transport code SPARTAN are described. The code is designed to deal with the very complex geometries encountered in lattice cell and fuel handling calculations, health physics, and shielding problems. Regions of the system being studied may be represented by simple shapes (spheres, cylinders, and so on) or by multinomial surfaces of any order, and many simple shapes may be combined to make up a complex layout. The geometry routines are designed to allow the program to carry out a number of tasks (such as sampling for a random point or tracking a path through several regions) in any order, so that the use of the routines is not restricted to a particular tracking or scoring method. Routines for reading, checking, and printing the data are included. (U.S.)
The geometry system of the three-dimensional Monte Carlo particle transport code SPARTAN
The geometry routines used in the general-purpose three-dimensional particle transport code SPARTAN are described. The code is designed to deal with the very complex geometries encountered in lattice cell and fuel handling calculations, health physics and shielding problems. Regions of the system being studied may be represented by simple shapes (spheres, cylinders etc) or by multinomial surfaces of any order, and many simple shapes may be combined to make up a complex layout. The geometry routines are designed to allow the program to carry out a number of tasks (such as sampling for a random point or tracking a path through several regions) in any order, so that the use of the routines is not restricted to a particular tracking of scoring method. Routines for reading, checking and printing the data are included. Details of the computational package are also included to indicate the way in which the generalised geometry capability of SPARTAN could be incorporated into other codes. (author)
Blind Decoding of Multiple Description Codes over OFDM Systems via Sequential Monte Carlo
Guo Dong
2005-01-01
Full Text Available We consider the problem of transmitting a continuous source through an OFDM system. Multiple description scalar quantization (MDSQ is applied to the source signal, resulting in two correlated source descriptions. The two descriptions are then OFDM modulated and transmitted through two parallel frequency-selective fading channels. At the receiver, a blind turbo receiver is developed for joint OFDM demodulation and MDSQ decoding. Transformation of the extrinsic information of the two descriptions are exchanged between each other to improve system performance. A blind soft-input soft-output OFDM detector is developed, which is based on the techniques of importance sampling and resampling. Such a detector is capable of exchanging the so-called extrinsic information with the other component in the above turbo receiver, and successively improving the overall receiver performance. Finally, we also treat channel-coded systems, and a novel blind turbo receiver is developed for joint demodulation, channel decoding, and MDSQ source decoding.
To avoid prohibitively long computation times, conventional Monte Carlo e-transport algorithms (e.g. EGS4, ETRAN, ITS) employ multiple scattering theories and ''condensed history'' methods to model e- transport. Although highly successful for many calculations, these techniques do not model backscatter very well, particularly for high-Z materials. In an attempt to correct for this shortcoming, we have extended the EGS4 Monte Carlo code to allow for the simulation of single elastic scattering. The single scattering method also allows quantities to be scored in submicrometer dimension geometries where the Moliere multiple scattering theory fails and the Goudsmit-Saunderson multiple scattering equations converge very slowly. Two single scattering schemes have been implemented: (i) Screened Rutherford cross sections which form the basis of Moliere's multiple scattering theory, (ii) Single scattering cross sections based upon phase-shift data. In this work we describe the implementation of single elastic scattering in the EGS4 Monte Carlo code system and employ it to verify the Moliere multiple scattering theory in its range of validity. We demonstrate that the Moliere multiple scattering formalism provides a good description of multiple scattering despite its use of a relatively crude cross section and that it may be employed with semi-quantitative accuracy in the plural scattering regime, where electron step-lengths are so short that only as few as five atoms participate in the angular deflection. However, the remaining differences of the Moliere distributions with the phase-shift data motivate the use of more accurate fundamental data, in particular, for applications involving high-Z elements. (orig.)
Criticality benchmarking of ANET Monte Carlo code
In this work the new Monte Carlo code ANET is tested on criticality calculations. ANET is developed based on the high energy physics code GEANT of CERN and aims at progressively satisfying several requirements regarding both simulations of GEN II/III reactors, as well as of innovative nuclear reactor designs such as the Accelerator Driven Systems (ADSs). Here ANET is applied on three different nuclear configurations, including a subcritical assembly, a Material Testing Reactor and the conceptual configuration of an ADS. In the first case, calculation of the effective multiplication factor (keff) are performed for the Training Nuclear Reactor of the Aristotle University of Thessaloniki, while in the second case keff is computed for the fresh fueled core of the Portuguese research reactor (RPJ) just after its conversion to Low Enriched Uranium, considering the control rods at the position that renders the reactor critical. In both cases ANET computations are compared with corresponding results obtained by three different well established codes, including both deterministic (XSDRNPM/CITATION) and Monte Carlo (TRIPOLI, MCNP). In the RPI case, keff computations are also compared with observations during the reactor core commissioning since the control rods are considered at criticality position. The above verification studies show ANET to produce reasonable results since they are satisfactorily compared with other models as well as with observations. For the third case (ADS), preliminary ANET computations of keff for various intensities of the proton beam are presented, showing also a reasonable code performance concerning both the order of magnitude and the relative variation of the computed parameter. (author)
The use of focused anti-scatter grids on digital radiographic systems with two-dimensional detectors produces acquisitions with a decreased scatter to primary ratio and thus improved contrast and resolution. Simulation software is of great interest in optimizing grid configuration according to a specific application. Classical simulators are based on complete detailed geometric descriptions of the grid. They are accurate but very time consuming since they use Monte Carlo code to simulate scatter within the high-frequency grids. We propose a new practical method which couples an analytical simulation of the grid interaction with a radiographic system simulation program. First, a two dimensional matrix of probability depending on the grid is created offline, in which the first dimension represents the angle of impact with respect to the normal to the grid lines and the other the energy of the photon. This matrix of probability is then used by the Monte Carlo simulation software in order to provide the final scattered flux image. To evaluate the gain of CPU time, we define the increasing factor as the increase of CPU time of the simulation with as opposed to without the grid. Increasing factors were calculated with the new model and with classical methods representing the grid with its CAD model as part of the object. With the new method, increasing factors are shorter by one to two orders of magnitude compared with the second one. These results were obtained with a difference in calculated scatter of less than five percent between the new and the classical method. (authors)
Coded aperture optimization using Monte Carlo simulations
Coded apertures using Uniformly Redundant Arrays (URA) have been unsuccessfully evaluated for two-dimensional and three-dimensional imaging in Nuclear Medicine. The images reconstructed from coded projections contain artifacts and suffer from poor spatial resolution in the longitudinal direction. We introduce a Maximum-Likelihood Expectation-Maximization (MLEM) algorithm for three-dimensional coded aperture imaging which uses a projection matrix calculated by Monte Carlo simulations. The aim of the algorithm is to reduce artifacts and improve the three-dimensional spatial resolution in the reconstructed images. Firstly, we present the validation of GATE (Geant4 Application for Emission Tomography) for Monte Carlo simulations of a coded mask installed on a clinical gamma camera. The coded mask modelling was validated by comparison between experimental and simulated data in terms of energy spectra, sensitivity and spatial resolution. In the second part of the study, we use the validated model to calculate the projection matrix with Monte Carlo simulations. A three-dimensional thyroid phantom study was performed to compare the performance of the three-dimensional MLEM reconstruction with conventional correlation method. The results indicate that the artifacts are reduced and three-dimensional spatial resolution is improved with the Monte Carlo-based MLEM reconstruction.
Successful vectorization - reactor physics Monte Carlo code
Most particle transport Monte Carlo codes in use today are based on the ''history-based'' algorithm, wherein one particle history at a time is simulated. Unfortunately, the ''history-based'' approach (present in all Monte Carlo codes until recent years) is inherently scalar and cannot be vectorized. In particular, the history-based algorithm cannot take advantage of vector architectures, which characterize the largest and fastest computers at the current time, vector supercomputers such as the Cray X/MP or IBM 3090/600. However, substantial progress has been made in recent years in developing and implementing a vectorized Monte Carlo algorithm. This algorithm follows portions of many particle histories at the same time and forms the basis for all successful vectorized Monte Carlo codes that are in use today. This paper describes the basic vectorized algorithm along with descriptions of several variations that have been developed by different researchers for specific applications. These applications have been mainly in the areas of neutron transport in nuclear reactor and shielding analysis and photon transport in fusion plasmas. The relative merits of the various approach schemes will be discussed and the present status of known vectorization efforts will be summarized along with available timing results, including results from the successful vectorization of 3-D general geometry, continuous energy Monte Carlo. (orig.)
Parallel processing Monte Carlo radiation transport codes
Issues related to distributed-memory multiprocessing as applied to Monte Carlo radiation transport are discussed. Measurements of communication overhead are presented for the radiation transport code MCNP which employs the communication software package PVM, and average efficiency curves are provided for a homogeneous virtual machine
Fonseca, T.C.F.; Bastos, F.M.; Figueiredo, M.T.T.; Souza, L.S.; Guimaraes, M.C.; Silva, C.R.E.; Mello, O.A.; Castelo e Silva, L.A.; Paixao, L., E-mail: tcff01@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Benavente, J.A.; Paiva, F.G. [Universidade Federal de Minas Gerais (PCTN/UFMG), Belo Horizonte, MG (Brazil). Curso de Pos-Graduacao em Ciencias e Tecnicas Nucleares
2015-07-01
Computational Monte Carlo (MC) codes have been used for simulation of nuclear installations mainly for internal monitoring of workers, the well known as Whole Body Counters (WBC). The main goal of this project was the modeling and simulation of the counting efficiency (CE) of a WBC system using three different MC codes: MCNPX, EGSnrc and VMC in-vivo. The simulations were performed for three different groups of analysts. The results shown differences between the three codes, as well as in the results obtained by the same code and modeled by different analysts. Moreover, all the results were also compared to the experimental results obtained in laboratory for meaning of validation and final comparison. In conclusion, it was possible to detect the influence on the results when the system is modeled by different analysts using the same MC code and in which MC code the results were best suited, when comparing to the experimental data result. (author)
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...
MORET: Version 4.B. A multigroup Monte Carlo criticality code
MORET 4 is a three dimensional multigroup Monte Carlo code which calculates the effective multiplication factor (keff) of any configurations more or less complex as well as reaction rates in the different volumes of the geometry and the leakage out of the system. MORET 4 is the Monte Carlo code of the APOLLO2-MORET 4 standard route of CRISTAL, the French criticality package. It is the most commonly used Monte Carlo code for French criticality calculations. During the last four years, the MORET 4 team has developed or improved the following major points: modernization of the geometry, implementation of perturbation algorithms, source distribution convergence, statistical detection of stationarity, unbiased variance estimation and creation of pre-processing and post-processing tools. The purpose of this paper is not only to present the new features of MORET but also to detail clearly the physical models and the mathematical methods used in the code. (author)
MCOR - Monte Carlo depletion code for reference LWR calculations
Research highlights: → Introduction of a reference Monte Carlo based depletion code with extended capabilities. → Verification and validation results for MCOR. → Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations
MCOR - Monte Carlo depletion code for reference LWR calculations
Puente Espel, Federico, E-mail: fup104@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Tippayakul, Chanatip, E-mail: cut110@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Ivanov, Kostadin, E-mail: kni1@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Misu, Stefan, E-mail: Stefan.Misu@areva.com [AREVA, AREVA NP GmbH, Erlangen (Germany)
2011-04-15
Research highlights: > Introduction of a reference Monte Carlo based depletion code with extended capabilities. > Verification and validation results for MCOR. > Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations. Additionally
This thesis has been performed in the framework of national reference setting-up for absorbed dose in water and high energy photon beam provided with the SATURNE-43 medical accelerator of the BNM-LPRI (acronym for National Bureau of Metrology and Primary standard laboratory of ionising radiation). The aim of this work has been to develop and validate different user codes, based on PENELOPE Monte Carlo code system, to determine the photon beam characteristics and calculate the correction factors of reference dosimeters such as Fricke dosimeters and graphite calorimeter. In the first step, the developed user codes have permitted the influence study of different components constituting the irradiation head. Variance reduction techniques have been used to reduce the calculation time. The phase space has been calculated for 6, 12 and 25 MV at the output surface level of the accelerator head, then used for calculating energy spectra and dose distributions in the reference water phantom. Results obtained have been compared with experimental measurements. The second step has been devoted to develop an user code allowing calculation correction factors associated with both BNM-LPRI's graphite and Fricke dosimeters thanks to a correlated sampling method starting with energy spectra obtained in the first step. Then the calculated correction factors have been compared with experimental and calculated results obtained with the Monte Carlo EGS4 code system. The good agreement, between experimental and calculated results, leads to validate simulations performed with the PENELOPE code system. (author)
Development of a New Monte Carlo reactor physics code
Leppänen, Jaakko
2007-01-01
Monte Carlo neutron transport codes are widely used in various reactor physics applications, traditionally related to criticality safety analyses, radiation shielding problems, detector modelling and validation of deterministic transport codes. The main advantage of the method is the capability to model geometry and interaction physics without major approximations. The disadvantage is that the modelling of complicated systems is very computing-intensive, which restricts the applications to so...
The MCNPX Monte Carlo Radiation Transport Code
MCNPX (Monte Carlo N-Particle eXtended) is a general-purpose Monte Carlo radiation transport code with three-dimensional geometry and continuous-energy transport of 34 particles and light ions. It contains flexible source and tally options, interactive graphics, and support for both sequential and multi-processing computer platforms. MCNPX is based on MCNP4c and has been upgraded to most MCNP5 capabilities. MCNP is a highly stable code tracking neutrons, photons and electrons, and using evaluated nuclear data libraries for low-energy interaction probabilities. MCNPX has extended this base to a comprehensive set of particles and light ions, with heavy ion transport in development. Models have been included to calculate interaction probabilities when libraries are not available. Recent additions focus on the time evolution of residual nuclei decay, allowing calculation of transmutation and delayed particle emission. MCNPX is now a code of great dynamic range, and the excellent neutronics capabilities allow new opportunities to simulate devices of interest to experimental particle physics, particularly calorimetry. This paper describes the capabilities of the current MCNPX version 2.6.C, and also discusses ongoing code development
THE MCNPX MONTE CARLO RADIATION TRANSPORT CODE
WATERS, LAURIE S. [Los Alamos National Laboratory; MCKINNEY, GREGG W. [Los Alamos National Laboratory; DURKEE, JOE W. [Los Alamos National Laboratory; FENSIN, MICHAEL L. [Los Alamos National Laboratory; JAMES, MICHAEL R. [Los Alamos National Laboratory; JOHNS, RUSSELL C. [Los Alamos National Laboratory; PELOWITZ, DENISE B. [Los Alamos National Laboratory
2007-01-10
MCNPX (Monte Carlo N-Particle eXtended) is a general-purpose Monte Carlo radiation transport code with three-dimensional geometry and continuous-energy transport of 34 particles and light ions. It contains flexible source and tally options, interactive graphics, and support for both sequential and multi-processing computer platforms. MCNPX is based on MCNP4B, and has been upgraded to most MCNP5 capabilities. MCNP is a highly stable code tracking neutrons, photons and electrons, and using evaluated nuclear data libraries for low-energy interaction probabilities. MCNPX has extended this base to a comprehensive set of particles and light ions, with heavy ion transport in development. Models have been included to calculate interaction probabilities when libraries are not available. Recent additions focus on the time evolution of residual nuclei decay, allowing calculation of transmutation and delayed particle emission. MCNPX is now a code of great dynamic range, and the excellent neutronics capabilities allow new opportunities to simulate devices of interest to experimental particle physics; particularly calorimetry. This paper describes the capabilities of the current MCNPX version 2.6.C, and also discusses ongoing code development.
SERPENT Monte Carlo reactor physics code
SERPENT is a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, developed at VTT Technical Research Centre of Finland since 2004. The code is specialized in lattice physics applications, but the universe-based geometry description allows transport simulation to be carried out in complicated three-dimensional geometries as well. The suggested applications of SERPENT include generation of homogenized multi-group constants for deterministic reactor simulator calculations, fuel cycle studies involving detailed assembly-level burnup calculations, validation of deterministic lattice transport codes, research reactor applications, educational purposes and demonstration of reactor physics phenomena. The Serpent code has been publicly distributed by the OECD/NEA Data Bank since May 2009 and RSICC in the U. S. since March 2010. The code is being used in some 35 organizations in 20 countries around the world. This paper presents an overview of the methods and capabilities of the Serpent code, with examples in the modelling of WWER-440 reactor physics. (Author)
Development of a New Monte Carlo reactor physics code
Monte Carlo neutron transport codes are widely used in various reactor physics applications, traditionally related to criticality safety analyses, radiation shielding problems, detector modelling and validation of deterministic transport codes. The main advantage of the method is the capability to model geometry and interaction physics without major approximations. The disadvantage is that the modelling of complicated systems is very computing-intensive, which restricts the applications to some extent. The importance of Monte Carlo calculation is likely to increase in the future, along with the development in computer capacities and parallel calculation. An interesting near-future application for the Monte Carlo method is the generation of input parameters for deterministic reactor simulator codes. These codes are used in coupled LWR full-core analyses and typically based on few-group nodal diffusion methods. The input data consists of homogenised few-group constants, presently generated using deterministic lattice transport codes. The task is becoming increasingly challenging, along with the development in nuclear technology. Calculations involving high-burnup fuels, advanced MOX technology and next-generation reactor systems are likely to cause problems in the future, if code development cannot keep up with the applications. A potential solution is the use of Monte Carlo based lattice transport codes, which brings all the advantages of the calculation method. So far there has been only a handful of studies on group constant generation using the Monte Carlo method, although the interest has clearly increased during the past few years. The homogenisation of reaction cross sections is simple and straightforward, and it can be carried out using any Monte Carlo code. Some of the parameters, however, require the use of special techniques that are usually not available in general-purpose codes. The main problem is the calculation of neutron diffusion coefficients, which
Rojas C, E.L.; Varon T, C.F.; Pedraza N, R. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: elrc@nuclear.inin.mx
2007-07-01
The treatment of the breast cancer at early stages is of vital importance. For that, most of the investigations are dedicated to the early detection of the suffering and their treatment. As investigation consequence and clinical practice, in 2002 it was developed in U.S.A. an irradiation system of high dose rate known as Mammosite. In this work we carry out dose calculations for a simplified Mammosite system with the Monte Carlo Penelope simulation code and MCNPX, varying the concentration of the contrast material that it is used in the one. (Author)
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
A semianalytic Monte Carlo code for modelling LIDAR measurements
Palazzi, Elisa; Kostadinov, Ivan; Petritoli, Andrea; Ravegnani, Fabrizio; Bortoli, Daniele; Masieri, Samuele; Premuda, Margherita; Giovanelli, Giorgio
2007-10-01
LIDAR (LIght Detection and Ranging) is an optical active remote sensing technology with many applications in atmospheric physics. Modelling of LIDAR measurements appears useful approach for evaluating the effects of various environmental variables and scenarios as well as of different measurement geometries and instrumental characteristics. In this regard a Monte Carlo simulation model can provide a reliable answer to these important requirements. A semianalytic Monte Carlo code for modelling LIDAR measurements has been developed at ISAC-CNR. The backscattered laser signal detected by the LIDAR system is calculated in the code taking into account the contributions due to the main atmospheric molecular constituents and aerosol particles through processes of single and multiple scattering. The contributions by molecular absorption, ground and clouds reflection are evaluated too. The code can perform simulations of both monostatic and bistatic LIDAR systems. To enhance the efficiency of the Monte Carlo simulation, analytical estimates and expected value calculations are performed. Artificial devices (such as forced collision, local forced collision, splitting and russian roulette) are moreover foreseen by the code, which can enable the user to drastically reduce the variance of the calculation.
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 UO2 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 UO2 and MOX single assembly models, based on the OECD-NEA/NRC PWR MOX/UO2 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
SPAMCART: a code for smoothed particle Monte Carlo radiative transfer
Lomax, O
2016-01-01
We present a code for generating synthetic SEDs and intensity maps from Smoothed Particle Hydrodynamics simulation snapshots. The code is based on the Lucy (1999) Monte Carlo Radiative Transfer method, i.e. it follows discrete luminosity packets, emitted from external and/or embedded sources, as they propagate through a density field, and then uses their trajectories to compute the radiative equilibrium temperature of the ambient dust. The density is not mapped onto a grid, and therefore the calculation is performed at exactly the same resolution as the hydrodynamics. We present two example calculations using this method. First, we demonstrate that the code strictly adheres to Kirchhoff's law of radiation. Second, we present synthetic intensity maps and spectra of an embedded protostellar multiple system. The algorithm uses data structures that are already constructed for other purposes in modern particle codes. It is therefore relatively simple to implement.
Parallel implementation of the Monte Carlo transport code EGS4 on the hypercube
Monte Carlo transport codes are commonly used in the study of particle interactions. The CALOR89 code system is a combination of several Monte Carlo transport and analysis programs. In order to produce good results, a typical Monte Carlo run will have to produce many particle histories. On a single processor computer, the transport calculation can take a huge amount of time. However, if the transport of particles were divided among several processors in a multiprocessor machine, the time can be drastically reduced
Proton interaction with an exposed object material needs to be modeled with account for three basic processes: electromagnetic stopping of protons in matter, multiple coulomb scattering and nuclear interactions. Just the last type of processes is the topic of this paper. Monte Carlo codes are often used to simulate high-energy particle interaction with matter. However, nuclear interaction models implemented in these codes are rather extensive and their use in treatment planning systems requires huge computational resources. We have selected the IThMC code for its ability to reproduce experiments which measure the distribution of the projected ranges of nuclear secondary particles generated by proton beams in a multi-layer Faraday cup. The multi-layer Faraday cup detectors measure charge rather than dose and allow distinguishing between electromagnetic and nuclear interactions. The event generator used in the IThMC code is faster, but less accurate than any other used in testing. Our model of nuclear reactions demonstrates quite good agreement with experiment in the context of their effect on the Bragg peak in therapeutic applications
Current work presents a new methodology which uses Serpent Monte-Carlo (MC) code for generating multi-group beginning-of-life (BOL) cross section (XS) database file that is compatible with PARCS 3D reactor core simulator and allows simulation of transients with the FAST code system. The applicability of the methodology was tested on European Sodium-cooled Fast Reactor (ESFR) design with an oxide fuel proposed by CEA (France). The k-effective, power peaking factors and safety parameters (such as Doppler constant, coolant density coefficient, fuel axial expansion coefficient, diagrid expansion coefficients and control rod worth) calculated by PARCS/TRACE were compared with the results of the Serpent MC code. The comparison indicates overall reasonable agreement between conceptually different (deterministic and stochastic) codes. The new development makes it in principle possible to use the Serpent MC code for cross section generation for the PARCS code to perform transient analyses for fast reactors. The advantages and limitations of this methodology are discussed in the paper. (author)
Mazurier, J
1999-05-28
This thesis has been performed in the framework of national reference setting-up for absorbed dose in water and high energy photon beam provided with the SATURNE-43 medical accelerator of the BNM-LPRI (acronym for National Bureau of Metrology and Primary standard laboratory of ionising radiation). The aim of this work has been to develop and validate different user codes, based on PENELOPE Monte Carlo code system, to determine the photon beam characteristics and calculate the correction factors of reference dosimeters such as Fricke dosimeters and graphite calorimeter. In the first step, the developed user codes have permitted the influence study of different components constituting the irradiation head. Variance reduction techniques have been used to reduce the calculation time. The phase space has been calculated for 6, 12 and 25 MV at the output surface level of the accelerator head, then used for calculating energy spectra and dose distributions in the reference water phantom. Results obtained have been compared with experimental measurements. The second step has been devoted to develop an user code allowing calculation correction factors associated with both BNM-LPRI's graphite and Fricke dosimeters thanks to a correlated sampling method starting with energy spectra obtained in the first step. Then the calculated correction factors have been compared with experimental and calculated results obtained with the Monte Carlo EGS4 code system. The good agreement, between experimental and calculated results, leads to validate simulations performed with the PENELOPE code system. (author)
Proton therapy Monte Carlo SRNA-VOX code
Ilić Radovan D.
2012-01-01
Full Text Available The most powerful feature of the Monte Carlo method is the possibility of simulating all individual particle interactions in three dimensions and performing numerical experiments with a preset error. These facts were the motivation behind the development of a general-purpose Monte Carlo SRNA program for proton transport simulation in technical systems described by standard geometrical forms (plane, sphere, cone, cylinder, cube. Some of the possible applications of the SRNA program are: (a a general code for proton transport modeling, (b design of accelerator-driven systems, (c simulation of proton scattering and degrading shapes and composition, (d research on proton detectors; and (e radiation protection at accelerator installations. This wide range of possible applications of the program demands the development of various versions of SRNA-VOX codes for proton transport modeling in voxelized geometries and has, finally, resulted in the ISTAR package for the calculation of deposited energy distribution in patients on the basis of CT data in radiotherapy. All of the said codes are capable of using 3-D proton sources with an arbitrary energy spectrum in an interval of 100 keV to 250 MeV.
Computed radiography simulation using the Monte Carlo code MCNPX
Simulating x-ray images has been of great interest in recent years as it makes possible an analysis of how x-ray images are affected owing to relevant operating parameters. In this paper, a procedure for simulating computed radiographic images using the Monte Carlo code MCNPX is proposed. The sensitivity curve of the BaFBr image plate detector as well as the characteristic noise of a 16-bit computed radiography system were considered during the methodology's development. The results obtained confirm that the proposed procedure for simulating computed radiographic images is satisfactory, as it allows obtaining results comparable with experimental data. (author)
Vectorization of continuous energy Monte Carlo code VIM
VIM is a continuous energy Monte Carlo code for criticality calculation. The random walk control system which uses combinatorial geometry system has been vectorized on FACOM VP-100. Vectorization has been done by the event bank method which controls simultaneous multiple particle's random walks, since behavior of neutron is independent. In vectorization of VIM code, we have two problems. One is a large overhead introduced by program modifications for vectorization. Another is a lowering of vector processing efficiency, since the vector length decreases with time according to the absorption and leakage of neutron and cut off of neutron for variance reduction. The average vector length during the random walks has been kept long by utilizing cross section library of single energy band and by reducing the number of the event banks. The performance ratio of vectorized version to the original one is 1.39 for the simple geometry and 1.13 for the complex geometry. (author)
Fast code for Monte Carlo simulations
A computer code to generate the dynamic evolution of the Ising model on a square lattice, following the Metropolis algorithm is presented. The computer time consumption is reduced by a factor of 8 when one compares our code with traditional multiple spin codes. The memory allocation size is also reduced by a factor of 4. The code is easily generalizable for other lattices and models. (author)
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.
MCNP-POLIMI v1.0, Monte Carlo N-Particle Transport Code System To Simulate Time-Analysis Quantities
1 - Description of program or function: MCNP is a general-purpose, continuous-energy, generalized geometry, time-dependent, coupled neutron-photon-electron Monte Carlo transport code system. Based on the Los Alamos National Laboratory code MCNP4C (formerly distributed as CCC-700), MCNP-PoliMi was developed to simulate time-analysis quantities. In particular, the code includes the correlation between neutron interaction and the corresponding photon production. Conversely to the technique adopted by standard MCNP, MCNP PoliMi samples secondary photons according to the neutron collision type. A post-processing code, i.e. the Matlab script 'postmain', is included and can be tailored to model specific detector characteristics. These features make MCNP-PoliMi a versatile tool to simulate particle interactions and detection processes. 2 - Methods: MCNP treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces. For neutrons, all reactions in a particular cross-section evaluation are accounted for. Both free gas and S(alpha, beta) thermal treatments are used. Criticality sources as well as fixed and surface sources are available. For photons, the code takes account of incoherent and coherent scattering with and without electron binding effects, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. A very general source and tally structure is available. The tallies have extensive statistical analysis of convergence. Rapid convergence is enabled by a wide variety of variance reduction methods. Energy ranges are 0-60 MeV for neutrons (data generally only available up to 20 MeV) and 1 keV - 1 GeV for photons and electrons. The MCNP-PoliMi code was developed to simulate each neutron-nucleus interaction as closely as possible. In particular, neutron interaction and
MOx benchmark calculations by deterministic and Monte Carlo codes
Highlights: ► MOx based depletion calculation. ► Methodology to create continuous energy pseudo cross section for lump of minor fission products. ► Mass inventory comparison between deterministic and Monte Carlo codes. ► Higher deviation was found for several isotopes. - Abstract: A depletion calculation benchmark devoted to MOx fuel is an ongoing objective of the OECD/NEA WPRS following the study of depletion calculation concerning UOx fuels. The objective of the proposed benchmark is to compare existing depletion calculations obtained with various codes and data libraries applied to fuel and back-end cycle configurations. In the present work the deterministic code NEWT/ORIGEN-S of the SCALE6 codes package and the Monte Carlo based code MONTEBURNS2.0 were used to calculate the masses of inventory isotopes. The methodology to apply the MONTEBURNS2.0 to this benchmark is also presented. Then the results from both code were compared.
A New Monte Carlo Neutron Transport Code at UNIST
Monte Carlo neutron transport code named MCS is under development at UNIST for the advanced reactor design and research purpose. This MC code can be used for fixed source calculation and criticality calculation. Continuous energy neutron cross section data and multi-group cross section data can be used for the MC calculation. This paper presents the overview of developed MC code and its calculation results. The real time fixed source calculation ability is also tested in this paper. The calculation results show good agreement with commercial code and experiment. A new Monte Carlo neutron transport code is being developed at UNIST. The MC codes are tested with several benchmark problems: ICSBEP, VENUS-2, and Hoogenboom-Martin benchmark. These benchmarks covers pin geometry to 3-dimensional whole core, and results shows good agreement with reference results
Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations
Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)
2015-09-15
A burn-up calculation of VVER's cores by Monte-Carlo code is complex process and requires large computational costs. This fact makes Monte-Carlo codes usage complicated for project and operating calculations. Previously prepared isotopic compositions are proposed to use for the Monte-Carlo code (MCU) calculations of different states of VVER's core with burnt fuel. Isotopic compositions are proposed to calculate by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by engineering codes (TVS-M, PERMAK-A). The multiplication factors and power distributions of FA and VVER with infinite height are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The MCU calculation data were compared with the data which were obtained by engineering codes.
Benchmarking Monte Carlo codes for criticality safety using subcritical measurements
Monte Carlo codes that are used for criticality safety evaluations are typically validated using critical experiments in which the neutron multiplication factor is unity. However, the conditions for most fissile material operations do not coincide to those of the critical experiments. This paper demonstrates that Monte Carlo methods and nuclear data can be validated using subcritical measurements whose conditions may coincide more closely to actual configurations of fissile material. (orig.)
Design of shielding of LILW containers by Monte Carlo codes
Accurate prediction of dose rates from containers with radioactive waste is becoming more important regarding more rigorous regulative in this area. The usual approach to the problem consists in combining numerical and measuring methods. In this paper a Monte Carlo calculations were used for calculating doses from a standard 200 liter drum which contains the intermediate level radioactive waste. Two different Monte Carlo codes were applied and compared, for the same combination of parameters. (author)
MCNP, a general Monte Carlo code for neutron and photon transport: a summary
The general-purpose Monte Carlo code MCNP can be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces
Progress on burnup calculation methods coupling Monte Carlo and depletion codes
Leszczynski, Francisco [Comision Nacional de Energia Atomica, San Carlos de Bariloche, RN (Argentina). Centro Atomico Bariloche]. E-mail: lesinki@cab.cnea.gob.ar
2005-07-01
Several methods of burnup calculations coupling Monte Carlo and depletion codes that were investigated and applied for the author last years are described. here. Some benchmark results and future possibilities are analyzed also. The methods are: depletion calculations at cell level with WIMS or other cell codes, and use of the resulting concentrations of fission products, poisons and actinides on Monte Carlo calculation for fixed burnup distributions obtained from diffusion codes; same as the first but using a method o coupling Monte Carlo (MCNP) and a depletion code (ORIGEN) at a cell level for obtaining the concentrations of nuclides, to be used on full reactor calculation with Monte Carlo code; and full calculation of the system with Monte Carlo and depletion codes, on several steps. All these methods were used for different problems for research reactors and some comparisons with experimental results of regular lattices were performed. On this work, a resume of all these works is presented and discussion of advantages and problems found are included. Also, a brief description of the methods adopted and MCQ system for coupling MCNP and ORIGEN codes is included. (author)
Highlights: • We investigated the main quantities determining ADS performance. • We calculated ADS performance such as neutron yield, neutron leakage spectra, heating and neutron and proton spectra in the target and in the beam window. • We used the MCNPX-2.7.0 Monte Carlo code for three-dimensional calculations. - Abstract: The MCNPX code offers options based on physics packages; the Bertini, ISABEL, INCL4 intra-nuclear models, and Dresner, ABLA evaporation–fission models and CEM2k cascade-exciton model. This study analyzes the main quantities determining ADS performance, such as neutron yield, neutron leakage spectra, heating and neutron and proton spectra in the target and in the beam window calculated by the MCNPX-2.5.0 Monte Carlo transport code, which is a combination of LAHET and MCNP codes. The results obtained by simulating different models cited above and implemented in MCNPX are compared with each other. The investigated system is composed of a natural lead cylindrical target and stainless steel (HT9) beam window. The target has been optimized to produce maximum number of neutrons with a radius of 20 cm and 70 cm of height. The target is bombarded with a high intensity linear accelerator by a 1 GeV, 1 mA proton beam. The protons are assumed uniformly distributed across the beam of radius 3 cm, and entering the target through a hole of 5.3 cm radius. The proton beam has an outer radius of 5.3 cm and an inner radius of 5.0 cm. The maximum value of the neutron flux in the target is observed on the axis ∼10 cm below the beam window, where the maximum difference between 7 different models is ∼15%. The total neutron leakage of the target calculated with the Bertini/ABLA is 1.83 × 1017 n/s, and is about 14% higher than the value calculated by the INCL4/Dresner (1.60 × 1017 n/s). Bertini/ABLA calculates top, bottom and side neutron leakage fractions as 20%, 2.3%, 77.6% of the total leakage, respectively, whereas, the calculated fractions are 18
Current status of the PSG Monte Carlo neutron transport code
PSG is a new Monte Carlo neutron transport code, developed at the Technical Research Centre of Finland (VTT). The code is mainly intended for fuel assembly-level reactor physics calculations, such as group constant generation for deterministic reactor simulator codes. This paper presents the current status of the project and the essential capabilities of the code. Although the main application of PSG is in lattice calculations, the geometry is not restricted in two dimensions. This paper presents the validation of PSG against the experimental results of the three-dimensional MOX fuelled VENUS-2 reactor dosimetry benchmark. (authors)
Vectorization techniques for neutron transport Monte Carlo codes
Four Monte Carlo codes, KENO IV, MORSE-DD, MCNP and VIM, have been vectorized already at JAERI Computing Center aiming at an increase in clculation performance, and speed-up ratios of vectorized codes to the original ones were found to be low values between 1.3 and 1.5. In this report the vectorization processes for these four codes are reviewed comprehensively, and methods of analysis for vectorization, modification of control structures of codes and debugging techniques are discussed. The reason for low speed-up ratios is also discussed. (author)
KAMCCO, a reactor physics Monte Carlo neutron transport code
KAMCCO is a 3-dimensional reactor Monte Carlo code for fast neutron physics problems. Two options are available for the solution of 1) the inhomogeneous time-dependent neutron transport equation (census time scheme), and 2) the homogeneous static neutron transport equation (generation cycle scheme). The user defines the desired output, e.g. estimates of reaction rates or neutron flux integrated over specified volumes in phase space and time intervals. Such primary quantities can be arbitrarily combined, also ratios of these quantities can be estimated with their errors. The Monte Carlo techniques are mostly analogue (exceptions: Importance sampling for collision processes, ELP/MELP, Russian roulette and splitting). Estimates are obtained from the collision and track length estimators. Elastic scattering takes into account first order anisotropy in the center of mass system. Inelastic scattering is processed via the evaporation model or via the excitation of discrete levels. For the calculation of cross sections, the energy is treated as a continuous variable. They are computed by a) linear interpolation, b) from optionally Doppler broadened single level Breit-Wigner resonances or c) from probability tables (in the region of statistically distributed resonances). (orig.)
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...
The Monte Carlo code TRAMO - Capabilities and instructions for application
The report is intended for readers familiar with the fundamentals of the Monte Carlo method. Those readers might be interested in learning about successful generalisations as well as new ideas for curbing the statistical errors involved. Another intention however is to explain the significant basic features of the multigroup Monte Carlo code TRAMO, including the required input, so that readers will be able to performing the required adjustments to the specific calculation technique and develop their own tools for performing their specific calculations. An indispensable code needed for such TRAMO applications is the TRAWEI Monte Carlo code which calculates he required weightings for applications of the variance reducing Weight Window Method; other codes required are those for generating the neutron cross-section data and the group data. The TRAMO code calculates, with given source distribution of neutrons in multigroup approximation, multigroup flux data, integrated group flux data, and dose values for given partial volumes and surfaces. There are further code versions for calculation of neutron and gamma fluxes, or criticality data, but these are not considered in the report. (orig./CB)
Taylor series development in the Monte Carlo code Tripoli-4
Mazzolo, Alain; Zoia, Andrea; Martin, Brunella
2014-06-01
Perturbation methods for one or several variables based on the Taylor series development up to the second order is presented for the collision estimator in the framework of the Monte Carlo code Tripoli-4. Comparisons with the correlated sampling method implemented in Tripoli-4 demonstrate the need of including the cross derivatives in the development.
Monte Carlo solver for UWB1 nuclear fuel depletion code
Highlights: • A new Monte Carlo solver was developed in order to speed-up depletion calculations. • For LWR model, UWB1 Monte Carlo solver is on average 10 times faster than MCNP6. • The UWB1 code will allow faster calculation analysis of BA parameters in fuel design. - Abstract: Recent nuclear reactor burnable absorber research tries to introduce new materials in the nuclear fuel. As a part of this effort, a fast computational tool is being developed for the advanced nuclear fuel. The first version of the newly developed UWB1 fast nuclear fuel depletion code significantly reduced calculation time by omitting the solution step for the Boltzmann transport equation. However, estimation of neutron multiplication factor during depletion was not sufficiently calculated. Therefore, at least one transport calculation for fuel depletion is necessary. This paper presents a new Monte Carlo solver that is implemented into the UWB1 code. The UWB1 Monte Carlo solver calculates neutron multiplication factor and neutron flux in the fuel for collapsed cross sections. Accuracy of the solver is supported by using current nuclear data stored in the ENDF/B-VII.1 library. Speed of the solver is the product of development focusing on minimization of CPU utilization at the expense of RAM demands. The UWB1 Monte Carlo solver is approximately 14 times faster than the MCNP6 reference code when one transport equation solution within fuel depletion is compared. Another speed-up can be achieved by employing advanced depletion scheme in the coupled transport and burnup equations. The resulting faster code will be used in optimization studies for ideal burnable absorber material selection where many various materials and concentrations will be evaluated
Acceleration of a Monte Carlo radiation transport code
Execution time for the Integrated TIGER Series (ITS) Monte Carlo radiation transport code has been reduced by careful re-coding of computationally intensive subroutines. Three test cases for the TIGER (1-D slab geometry), CYLTRAN (2-D cylindrical geometry), and ACCEPT (3-D arbitrary geometry) codes were identified and used to benchmark and profile program execution. Based upon these results, sixteen top time-consuming subroutines were examined and nine of them modified to accelerate computations with equivalent numerical output to the original. The results obtained via this study indicate that speedup factors of 1.90 for the TIGER code, 1.67 for the CYLTRAN code, and 1.11 for the ACCEPT code are achievable. copyright 1996 American Institute of Physics
MCNP: a general Monte Carlo code for neutron and photon transport. Version 3A. Revision 2
This manual is a practical guide for the use of our general-purpose Monte Carlo code MCNP. The first chapter is a primer for the novice user. The second chapter describes the mathematics, data, physics, and Monte Carlo simulation found in MCNP. This discussion is not meant to be exhaustive - details of the particular techniques and of the Monte Carlo method itself will have to be found elsewhere. The third chapter shows the user how to prepare input for the code. The fourth chapter contains several examples, and the fifth chapter explains the output. The appendices show how to use MCNP on particular computer systems at the Los Alamos National Laboratory and also give details about some of the code internals that those who wish to modify the code may find useful. 57 refs
Study on random number generator in Monte Carlo code
The Monte Carlo code uses a sequence of pseudo-random numbers with a random number generator (RNG) to simulate particle histories. A pseudo-random number has its own period depending on its generation method and the period is desired to be long enough not to exceed the period during one Monte Carlo calculation to ensure the correctness especially for a standard deviation of results. The linear congruential generator (LCG) is widely used as Monte Carlo RNG and the period of LCG is not so long by considering the increasing rate of simulation histories in a Monte Carlo calculation according to the remarkable enhancement of computer performance. Recently, many kinds of RNG have been developed and some of their features are better than those of LCG. In this study, we investigate the appropriate RNG in a Monte Carlo code as an alternative to LCG especially for the case of enormous histories. It is found that xorshift has desirable features compared with LCG, and xorshift has a larger period, a comparable speed to generate random numbers, a better randomness, and good applicability to parallel calculation. (author)
In this paper, the EGS4 Monte Carlo simulation system was used to develop a computer code for a study of the response of a NaI(Tl) detector exposed to γ-rays with energies below 300 keV. This study comprised registration of the spectra of the incident rays and determination of the photo peaks. In addition, the probability of the K X-ray escape from a NaI(Tl) crystal and its dependence on the detector shape and volume were considered. The results of the Monte Carlo simulation are in good agreement with the experimental data (the estimated discrepancy is below 5%). This demonstrates a high efficiency of the used simulation code in quantifying the physical parameters that are difficult to evaluate by experimental methods
Burnup calculations of TR-2 Research Reactor with Monteburns Monte Carlo Code
Full text: In this study, some neutronic calculations of first and second core cycles of 5 MW pool type TR-2 Research Reactor have been performed using Multi-Step Monte Carlo Burnup Code System MONTEBURNS and the results were compared with the values of experiments and other codes. Time dependent keff distribution and burnup ratios belong to first and second core cycles of TR-2 Research Reactor were compared and quite good consistence in the results were observed. After modeling the first and second core cycles of TR-2 with MCNP5 Monte Carlo code, MCNP5 used in MONTEBURNS code has been parallelized in 8 HP ProLiant BL680C G5 systems with 4 quad-core Intel Xeon E7330 CPU, utilizing the MPI parallel protocol and simulations were performed on the 128 cores Linux parallel computing machine system. The computation time was reduced by parallelization of MONTEBURNS which uses MCNP in many steps. (authors)
MCAM 5: an advanced interface program for multiple Monte Carlo Codes
The Automatic Modeling Program for Neutronics and Radiation Transport Simulation (MCAM) developed in China, is an advanced interface program between CAD (Computer Aided Design) systems and Monte Carlo (MC) codes. It can significantly reduce the manpower and enhance reliability for constructing MC codes input of complex systems. The latest version MCAM 4.8 was a mature and efficient version which was benchmarked with ITER benchmark model and has been used by hundreds of institutes in more than 40 countries all over the world. It can deal with MCNP and TRIPOLI models. The main function of MCAM is to convert geometries in CAD systems to geometries in MC codes input files. The MCAM version 5.2 is going to be released with added capabilities to support SuperMC, Geant4 and FLUKA Monte Carlo codes
TRIPOLI-3: a neutron/photon Monte Carlo transport code
The present version of TRIPOLI-3 solves the transport equation for coupled neutron and gamma ray problems in three dimensional geometries by using the Monte Carlo method. This code is devoted both to shielding and criticality problems. The most important feature for particle transport equation solving is the fine treatment of the physical phenomena and sophisticated biasing technics useful for deep penetrations. The code is used either for shielding design studies or for reference and benchmark to validate cross sections. Neutronic studies are essentially cell or small core calculations and criticality problems. TRIPOLI-3 has been used as reference method, for example, for resonance self shielding qualification. (orig.)
Burnup calculation methodology in the serpent 2 Monte Carlo code
This paper presents two topics related to the burnup calculation capabilities in the Serpent 2 Monte Carlo code: advanced time-integration methods and improved memory management, accomplished by the use of different optimization modes. The development of the introduced methods is an important part of re-writing the Serpent source code, carried out for the purpose of extending the burnup calculation capabilities from 2D assembly-level calculations to large 3D reactor-scale problems. The progress is demonstrated by repeating a PWR test case, originally carried out in 2009 for the validation of the newly-implemented burnup calculation routines in Serpent 1. (authors)
On the inner workings of Monte Carlo codes
Dubbeldam, D.; Torres Knoop, A.; Walton, K.S.
2013-01-01
We review state-of-the-art Monte Carlo (MC) techniques for computing fluid coexistence properties (Gibbs simulations) and adsorption simulations in nanoporous materials such as zeolites and metal-organic frameworks. Conventional MC is discussed and compared to advanced techniques such as reactive MC, configurational-bias Monte Carlo and continuous fractional MC. The latter technique overcomes the problem of low insertion probabilities in open systems. Other modern methods are (hyper-)parallel...
Proton therapy Monte Carlo SRNA-VOX code
Ilić Radovan D.
2012-01-01
The most powerful feature of the Monte Carlo method is the possibility of simulating all individual particle interactions in three dimensions and performing numerical experiments with a preset error. These facts were the motivation behind the development of a general-purpose Monte Carlo SRNA program for proton transport simulation in technical systems described by standard geometrical forms (plane, sphere, cone, cylinder, cube). Some of the possible applications of the SRNA program are:...
Full-core pin-power calculations using Monte Carlo codes
Pin wise calculations of core power distribution have been performed for a criticality mock up installation that models a WWER-1000 reactor. Two Monte Carlo codes have been applied for solving of this problem: the MCNP4B code and the KENO-VI code from the SCALE 4.4 system. The codes use different kinds of neutron cross section data: pointwise continuous-energy ENDF/B-VI data and multigroup ENDF/B-V data. Comparisons of calculated results show that the MCNP4B and KENO-VI results are in good agreement. (authors)
Verification of Monte Carlo transport codes by activation experiments
With the increasing energies and intensities of heavy-ion accelerator facilities, the problem of an excessive activation of the accelerator components caused by beam losses becomes more and more important. Numerical experiments using Monte Carlo transport codes are performed in order to assess the levels of activation. The heavy-ion versions of the codes were released approximately a decade ago, therefore the verification is needed to be sure that they give reasonable results. Present work is focused on obtaining the experimental data on activation of the targets by heavy-ion beams. Several experiments were performed at GSI Helmholtzzentrum fuer Schwerionenforschung. The interaction of nitrogen, argon and uranium beams with aluminum targets, as well as interaction of nitrogen and argon beams with copper targets was studied. After the irradiation of the targets by different ion beams from the SIS18 synchrotron at GSI, the γ-spectroscopy analysis was done: the γ-spectra of the residual activity were measured, the radioactive nuclides were identified, their amount and depth distribution were detected. The obtained experimental results were compared with the results of the Monte Carlo simulations using FLUKA, MARS and SHIELD. The discrepancies and agreements between experiment and simulations are pointed out. The origin of discrepancies is discussed. Obtained results allow for a better verification of the Monte Carlo transport codes, and also provide information for their further development. The necessity of the activation studies for accelerator applications is discussed. The limits of applicability of the heavy-ion beam-loss criteria were studied using the FLUKA code. FLUKA-simulations were done to determine the most preferable from the radiation protection point of view materials for use in accelerator components.
Longitudinal development of extensive air showers: hybrid code SENECA and full Monte Carlo
Ortiz, J A; De Souza, V; Ortiz, Jeferson A.; Tanco, Gustavo Medina
2004-01-01
New experiments, exploring the ultra-high energy tail of the cosmic ray spectrum with unprecedented detail, are exerting a severe pressure on extensive air hower modeling. Detailed fast codes are in need in order to extract and understand the richness of information now available. Some hybrid simulation codes have been proposed recently to this effect (e.g., the combination of the traditional Monte Carlo scheme and system of cascade equations or pre-simulated air showers). In this context, we explore the potential of SENECA, an efficient hybrid tridimensional simulation code, as a valid practical alternative to full Monte Carlo simulations of extensive air showers generated by ultra-high energy cosmic rays. We extensively compare hybrid method with the traditional, but time consuming, full Monte Carlo code CORSIKA which is the de facto standard in the field. The hybrid scheme of the SENECA code is based on the simulation of each particle with the traditional Monte Carlo method at two steps of the shower devel...
Purpose: Aim of this study was to optimize the magnetic field strengths of two quadrupole magnets in a particle therapy facility in order to obtain a beam quality suitable for spot beam scanning. Methods: The particle transport through an ion-optic system of a particle therapy facility consisting of the beam tube, two quadrupole magnets and a beam monitor system was calculated with the help of Matlab by using matrices that solve the equation of motion of a charged particle in a magnetic field and field-free region, respectively. The magnetic field strengths were optimized in order to obtain a circular and thin beam spot at the iso-center of the therapy facility. These optimized field strengths were subsequently transferred to the Monte-Carlo code FLUKA and the transport of 80 MeV/u C12-ions through this ion-optic system was calculated by using a user-routine to implement magnetic fields. The fluence along the beam-axis and at the iso-center was evaluated. Results: The magnetic field strengths could be optimized by using Matlab and transferred to the Monte-Carlo code FLUKA. The implementation via a user-routine was successful. Analyzing the fluence-pattern along the beam-axis the characteristic focusing and de-focusing effects of the quadrupole magnets could be reproduced. Furthermore the beam spot at the iso-center was circular and significantly thinner compared to an unfocused beam. Conclusion: In this study a Matlab tool was developed to optimize magnetic field strengths for an ion-optic system consisting of two quadrupole magnets as part of a particle therapy facility. These magnetic field strengths could subsequently be transferred to and implemented in the Monte-Carlo code FLUKA to simulate the particle transport through this optimized ion-optic system
Nanodosimetric verification in proton therapy: Monte Carlo Codes Comparison
Full text: Nanodosimetry strives to develop a novel dosimetry concept suitable for advanced modalities of cancer radiotherapy, such as proton therapy. This project aims to evaluate the plausibility of the physical models implemented in the Geant4 Very Low Energy (Geant4-DNA) extensions by comparing nanodosimetric quantities calculated with Geant4-DNA and the PTB Monte Carlo track structure code. Nanodosimetric track structure parameters were calculated for cylindrical targets representing DNA and nucleosome segments and converted into the probability of producing a DSB using the model proposed by Garty et al. [1]. Monoenergetic protons and electrons of energies typical for 6-electron spectra were considered as primary particles. Good agreement was found between the two codes for electrons of energies above 200 eV. Below this energy Geant4-DNA produced slightly higher numbers of ionisations in the sensitive volumes and higher probabilities for DSB formation. For protons, Geant4-DNA also gave higher numbers of ionisations and DSB probabilities, particularly in the low energy range, while a satisfactory agreement was found for energies higher than I MeV. Comparing two codes can be useful as any observed divergence in results between the two codes provides valuable information as to where further consideration of the underlying physical models used in each code may be required. Consistently it was seen that the largest difference between the codes was in the low energy ranges for each particle type. (author)
Modelling photon transport in non-uniform media for SPECT with a vectorized Monte Carlo code.
Smith, M F
1993-10-01
A vectorized Monte Carlo code has been developed for modelling photon transport in non-uniform media for single-photon-emission computed tomography (SPECT). The code is designed to compute photon detection kernels, which are used to build system matrices for simulating SPECT projection data acquisition and for use in matrix-based image reconstruction. Non-uniform attenuating and scattering regions are constructed from simple three-dimensional geometric shapes, in which the density and mass attenuation coefficients are individually specified. On a Stellar GS1000 computer, Monte Carlo simulations are performed between 1.6 and 2.0 times faster when the vector processor is utilized than when computations are performed in scalar mode. Projection data acquired with a clinical SPECT gamma camera for a line source in a non-uniform thorax phantom are well modelled by Monte Carlo simulations. The vectorized Monte Carlo code was used to stimulate a 99Tcm SPECT myocardial perfusion study, and compensations for non-uniform attenuation and the detection of scattered photons improve activity estimation. The speed increase due to vectorization makes Monte Carlo simulation more attractive as a tool for modelling photon transport in non-uniform media for SPECT. PMID:8248288
Adjoint Monte Carlo techniques and codes for organ dose calculations
Adjoint Monte Carlo simulations can be effectively used for the estimation of doses in small targets when the sources are extended in large volumes or surfaces. The main features of two computer codes for calculating doses at free points or in organs of an anthropomorphic phantom are described. In the first program (REBEL-3) natural gamma-emitting sources are contained in the walls of a dwelling room; in the second one (POKER-CAMP) the user can specify arbitrary gamma sources with different spatial distributions in the environment: in (or on the surface of) the ground and in the air. 3 figures
ALEPH 1.1.2: A Monte Carlo burn-up code
In the last 40 years, Monte Carlo particle transport has been applied to a multitude of problems such as shielding and medical applications, to various types of nuclear reactors, . . . The success of the Monte Carlo method is mainly based on its broad application area, on its ability to handle nuclear data not only in its most basic but also most complex form (namely continuous energy cross sections, complex interaction laws, detailed energy-angle correlations, multi-particle physics, . . . ), on its capability of modeling geometries from simple 1D to complex 3D, . . . There is also a current trend in Monte Carlo applications toward high detail 3D calculations (for instance voxel-based medical applications), something for which deterministic codes are neither suited nor performant as to computational time and precision. Apart from all these fields where Monte Carlo particle transport has been applied successfully, there is at least one area where Monte Carlo has had limited success, namely burn-up and activation calculations where the time parameter is added to the problem. The concept of Monte Carlo burn-up consists of coupling a Monte Carlo code to a burn-up module to improve the accuracy of depletion and activation calculations. For every time step the Monte Carlo code will provide reaction rates to the burn-up module which will return new material compositions to the Monte Carlo code. So if static Monte Carlo particle transport is slow, then Monte Carlo particle transport with burn-up will be even slower as calculations have to be performed for every time step in the problem. The computational issues to perform accurate Monte Carlo calculations are however continuously reduced due to improvements made in the basic Monte Carlo algorithms, due to the development of variance reduction techniques and due to developments in computer architecture (more powerful processors, the so-called brute force approach through parallel processors and networked systems
MCNP: a general Monte Carlo code for neutron and photon transport
MCNP is a very general Monte Carlo neutron photon transport code system with approximately 250 person years of Group X-6 code development invested. It is extremely portable, user-oriented, and a true production code as it is used about 60 Cray hours per month by about 150 Los Alamos users. It has as its data base the best cross-section evaluations available. MCNP contains state-of-the-art traditional and adaptive Monte Carlo techniques to be applied to the solution of an ever-increasing number of problems. Excellent user-oriented documentation is available for all facets of the MCNP code system. Many useful and important variants of MCNP exist for special applications. The Radiation Shielding Information Center (RSIC) in Oak Ridge, Tennessee is the contact point for worldwide MCNP code and documentation distribution. A much improved MCNP Version 3A will be available in the fall of 1985, along with new and improved documentation. Future directions in MCNP development will change the meaning of MCNP to Monte Carlo N Particle where N particle varieties will be transported
TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files
Cullen, D E
1998-11-22
TART98 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input preparation, running Monte Carlo calculations, and analysis of output results. TART98 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART98 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART98 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART98 and its data files.
Parallel computing by Monte Carlo codes MVP/GMVP
General-purpose Monte Carlo codes MVP/GMVP are well-vectorized and thus enable us to perform high-speed Monte Carlo calculations. In order to achieve more speedups, we parallelized the codes on the different types of parallel computing platforms or by using a standard parallelization library MPI. The platforms used for benchmark calculations are a distributed-memory vector-parallel computer Fujitsu VPP500, a distributed-memory massively parallel computer Intel paragon and a distributed-memory scalar-parallel computer Hitachi SR2201, IBM SP2. As mentioned generally, linear speedup could be obtained for large-scale problems but parallelization efficiency decreased as the batch size per a processing element(PE) was smaller. It was also found that the statistical uncertainty for assembly powers was less than 0.1% by the PWR full-core calculation with more than 10 million histories and it took about 1.5 hours by massively parallel computing. (author)
Simulating fast transients with fuel behavior feedback using the Serpent 2 Monte Carlo code
Simulating transients with reactivity feedback effects using Monte Carlo neutron transport codes can be used for validating deterministic transient codes or estimating for example the total deposited energy in a fuel rod following a known reactivity insertion in the system. Recent increases in computational power as well as developments in calculation methodology makes it possible to obtain a coupled solution for several aspects of the multi-physics problem in a single calculation. This paper describes the different methods implemented in Serpent 2 Monte Carlo code that enable it to model fast transients with fuel behavior feedback. The capability is demonstrated in a prompt critical pin-cell case, where the transient is shut down by the negative reactivity from rising fuel temperature. (author)
A comparison between the Monte Carlo radiation transport codes MCNP and MCBEND
Sawamura, Hidenori; Nishimura, Kazuya [Computer Software Development Co., Ltd., Tokyo (Japan)
2001-01-01
In Japan, almost of all radiation analysts are using the MCNP code and MVP code on there studies. But these codes have not had automatic variance reduction. MCBEND code made by UKAEA have automatic variance reduction. And, MCBEND code is user friendly more than other Monte Carlo Radiation Transport Codes. Our company was first introduced MCBEND code in Japan. Therefore, we compared with MCBEND code and MCNP code about functions and production capacity. (author)
Platt, M. E.; Lewis, E. E.; Boehm, F.
1991-01-01
A Monte Carlo Fortran computer program was developed that uses two variance reduction techniques for computing system reliability applicable to solving very large highly reliable fault-tolerant systems. The program is consistent with the hybrid automated reliability predictor (HARP) code which employs behavioral decomposition and complex fault-error handling models. This new capability is called MC-HARP which efficiently solves reliability models with non-constant failures rates (Weibull). Common mode failure modeling is also a specialty.
Criticality qualification of a new Monte Carlo code for reactor core analysis
In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal-hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the 'Accelerator part' of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the 'Reactor part' of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.
Criticality qualification of a new Monte Carlo code for reactor core analysis
Catsaros, N. [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece); Gaveau, B. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Jaekel, M. [Laboratoire de Physique Theorique, Ecole Normale Superieure, 24 rue Lhomond, 75231 Paris (France); Maillard, J. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); CNRS-IDRIS, Bt 506, BP167, 91403 Orsay (France); CNRS-IN2P3, 3 rue Michel Ange, 75794 Paris (France); Maurel, G. [Faculte de Medecine, Universite Paris VI, 27 rue de Chaligny, 75012 Paris (France); MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Savva, P., E-mail: savvapan@ipta.demokritos.g [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece); Silva, J. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Varvayanni, M.; Zisis, Th. [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece)
2009-11-15
In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal-hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the 'Accelerator part' of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the 'Reactor part' of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.
Monte Carlo simulation of medical linear accelerator using primo code
The use of monte Carlo simulation has become very important in the medical field and especially in calculation in radiotherapy. Various Monte Carlo codes were developed simulating interactions of particles and photons with matter. One of these codes is PRIMO that performs simulation of radiation transport from the primary electron source of a linac to estimate the absorbed dose in a water phantom or computerized tomography (CT). PRIMO is based on Penelope Monte Carlo code. Measurements of 6 MV photon beam PDD and profile were done for Elekta precise linear accelerator at Radiation and Isotopes Center Khartoum using computerized Blue water phantom and CC13 Ionization Chamber. accept Software was used to control the phantom to measure and verify dose distribution. Elektalinac from the list of available linacs in PRIMO was tuned to model Elekta precise linear accelerator. Beam parameter of 6.0 MeV initial electron energy, 0.20 MeV FWHM, and 0.20 cm focal spot FWHM were used, and an error of 4% between calculated and measured curves was found. The buildup region Z max was 1.40 cm and homogenous profile in cross line and in line were acquired. A number of studies were done to verily the model usability one of them is the effect of the number of histories on accuracy of the simulation and the resulted profile for the same beam parameters. The effect was noticeable and inaccuracies in the profile were reduced by increasing the number of histories. Another study was the effect of Side-step errors on the calculated dose which was compared with the measured dose for the same setting.It was in range of 2% for 5 cm shift, but it was higher in the calculated dose because of the small difference between the tuned model and measured dose curves. Future developments include simulating asymmetrical fields, calculating the dose distribution in computerized tomographic (CT) volume, studying the effect of beam modifiers on beam profile for both electron and photon beams.(Author)
The development of depletion program coupled with Monte Carlo computer code
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 MCNPREBUS linkage system code. The results show good agreement between calculated data of the MCDL code and reference codes. (author)
Verification of Monte Carlo transport codes FLUKA, Mars and Shield
The present study is a continuation of the project 'Verification of Monte Carlo Transport Codes' which is running at GSI as a part of activation studies of FAIR relevant materials. It includes two parts: verification of stopping modules of FLUKA, MARS and SHIELD-A (with ATIMA stopping module) and verification of their isotope production modules. The first part is based on the measurements of energy deposition function of uranium ions in copper and stainless steel. The irradiation was done at 500 MeV/u and 950 MeV/u, the experiment was held at GSI from September 2004 until May 2005. The second part is based on gamma-activation studies of an aluminium target irradiated with an argon beam of 500 MeV/u in August 2009. Experimental depth profiling of the residual activity of the target is compared with the simulations. (authors)
The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged
Parallel Grand Canonical Monte Carlo (ParaGrandMC) Simulation Code
Yamakov, Vesselin I.
2016-01-01
This report provides an overview of the Parallel Grand Canonical Monte Carlo (ParaGrandMC) simulation code. This is a highly scalable parallel FORTRAN code for simulating the thermodynamic evolution of metal alloy systems at the atomic level, and predicting the thermodynamic state, phase diagram, chemical composition and mechanical properties. The code is designed to simulate multi-component alloy systems, predict solid-state phase transformations such as austenite-martensite transformations, precipitate formation, recrystallization, capillary effects at interfaces, surface absorption, etc., which can aid the design of novel metallic alloys. While the software is mainly tailored for modeling metal alloys, it can also be used for other types of solid-state systems, and to some degree for liquid or gaseous systems, including multiphase systems forming solid-liquid-gas interfaces.
The Physical Models and Statistical Procedures Used in the RACER Monte Carlo Code
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
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...
Monte Carlo simulation in UWB1 depletion code
UWB1 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
The Monte Carlo code MCSHAPE: Main features and recent developments
MCSHAPE is a general purpose Monte Carlo code developed at the University of Bologna to simulate the diffusion of X- and gamma-ray photons with the special feature of describing the full evolution of the photon polarization state along the interactions with the target. The prevailing photon–matter interactions in the energy range 1–1000 keV, Compton and Rayleigh scattering and photoelectric effect, are considered. All the parameters that characterize the photon transport can be suitably defined: (i) the source intensity, (ii) its full polarization state as a function of energy, (iii) the number of collisions, and (iv) the energy interval and resolution of the simulation. It is possible to visualize the results for selected groups of interactions. MCSHAPE simulates the propagation in heterogeneous media of polarized photons (from synchrotron sources) or of partially polarized sources (from X-ray tubes). In this paper, the main features of MCSHAPE are illustrated with some examples and a comparison with experimental data. - Highlights: • MCSHAPE is an MC code for the simulation of the diffusion of photons in the matter. • It includes the proper description of the evolution of the photon polarization state. • The polarization state is described by means of the Stokes vector, I, Q, U, V. • MCSHAPE includes the computation of the detector influence in the measured spectrum. • MCSHAPE features are illustrated with examples and comparison with experiments
Development and verification of Monte Carlo burnup calculation system
Monte Carlo burnup calculation code system has been developed to evaluate accurate various quantities required in the backend field. From the Actinide Research in a Nuclear Element (ARIANE) program, by using, the measured nuclide compositions of fuel rods in the fuel assemblies irradiated in the commercial Netherlands BWR, the analyses have been performed for the code system verification. The code system developed in this paper has been verified through analysis for MOX and UO2 fuel rods. This system enables to reduce large margin assumed in the present criticality analysis for LWR spent fuels. (J.P.N.)
Monte Carlo N Particle code - Dose distribution of clinical electron beams in inhomogeneous phantoms
H A Nedaie; Mosleh-Shirazi, M. A.; Allahverdi, M.
2013-01-01
Electron dose distributions calculated using the currently available analytical methods can be associated with large uncertainties. The Monte Carlo method is the most accurate method for dose calculation in electron beams. Most of the clinical electron beam simulation studies have been performed using non- MCNP [Monte Carlo N Particle] codes. Given the differences between Monte Carlo codes, this work aims to evaluate the accuracy of MCNP4C-simulated electron dose distributions in a homogenous...
ITS - The integrated TIGER series of coupled electron/photon Monte Carlo transport codes
The TIGER series of time-independent coupled electron/photon Monte Carlo transport codes is a group of multimaterial, multidimensional codes designed to provide a state-of-the-art description of the production and transport of the electron/photon cascade. The codes follow both electrons and photons from 1.0 GeV down to 1.0 keV, and the user has the option of combining the collisional transport with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence. Source particles can be either electrons or photons. The most important output data are (a) charge and energy deposition profiles, (b) integral and differential escape coefficients for both electrons and photons, (c) differential electron and photon flux, and (d) pulse-height distributions for selected regions of the problem geometry. The base codes of the series differ from one another primarily in their dimensionality and geometric modeling. They include (a) a one-dimensional multilayer code, (b) a code that describes the transport in two-dimensional axisymmetric cylindrical material geometries with a fully three-dimensional description of particle trajectories, and (c) a general three-dimensional transport code which employs a combinatorial geometry scheme. These base codes were designed primarily for describing radiation transport for those situations in which the detailed atomic structure of the transport medium is not important. For some applications, it is desirable to have a more detailed model of the low energy transport. The system includes three additional codes that contain a more elaborate ionization/relaxation model than the base codes. Finally, the system includes two codes that combine the collisional transport of the multidimensional base codes with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence
Review of the Monte Carlo and deterministic codes in radiation protection and dosimetry
Modelling a physical system can be carried out either stochastically or deterministically. An example of the former method is the Monte Carlo technique, in which statistically approximate methods are applied to exact models. No transport equation is solved as individual particles are simulated and some specific aspect (tally) of their average behaviour is recorded. The average behaviour of the physical system is then inferred using the central limit theorem. In contrast, deterministic codes use mathematically exact methods that are applied to approximate models to solve the transport equation for the average particle behaviour. The physical system is subdivided in boxes in the phase-space system and particles are followed from one box to the next. The smaller the boxes the better the approximations become. Although the Monte Carlo method has been used for centuries, its more recent manifestation has really emerged from the Manhattan project of the Word War II. Its invention is thought to be mainly due to Metropolis, Ulah (through his interest in poker), Fermi, von Neuman and Richtmeyer. Over the last 20 years or so, the Monte Carlo technique has become a powerful tool in radiation transport. This is due to users taking full advantage of richer cross section data, more powerful computers and Monte Carlo techniques for radiation transport, with high quality physics and better known source spectra. This method is a common sense approach to radiation transport and its success and popularity is quite often also due to necessity, because measurements are not always possible or affordable. In the Monte Carlo method, which is inherently realistic because nature is statistical, a more detailed physics is made possible by isolation of events while rather elaborate geometries can be modelled. Provided that the physics is correct, a simulation is exactly analogous to an experimenter counting particles. In contrast to the deterministic approach, however, a disadvantage of the
Method of tallying adjoint fluence and calculating kinetics parameters in Monte Carlo codes
A method of using iterated fission probability to estimate the adjoint fluence during particles simulation, and using it as the weighting function to calculate kinetics parameters βeff and A in Monte Carlo codes, was introduced in this paper. Implements of this method in continuous energy Monte Carlo code MCNP and multi-group Monte Carlo code MCMG are both elaborated. Verification results show that, with regardless additional computing cost, using this method, the adjoint fluence accounted by MCMG matches well with the result computed by ANISN, and the kinetics parameters calculated by MCNP agree very well with benchmarks. This method is proved to be reliable, and the function of calculating kinetics parameters in Monte Carlo codes is carried out effectively, which could be the basement for Monte Carlo codes' utility in the analysis of nuclear reactors' transient behavior. (authors)
Implementation of a Monte Carlo based inverse planning model for clinical IMRT with MCNP code
He, Tongming Tony
In IMRT inverse planning, inaccurate dose calculations and limitations in optimization algorithms introduce both systematic and convergence errors to treatment plans. The goal of this work is to practically implement a Monte Carlo based inverse planning model for clinical IMRT. The intention is to minimize both types of error in inverse planning and obtain treatment plans with better clinical accuracy than non-Monte Carlo based systems. The strategy is to calculate the dose matrices of small beamlets by using a Monte Carlo based method. Optimization of beamlet intensities is followed based on the calculated dose data using an optimization algorithm that is capable of escape from local minima and prevents possible pre-mature convergence. The MCNP 4B Monte Carlo code is improved to perform fast particle transport and dose tallying in lattice cells by adopting a selective transport and tallying algorithm. Efficient dose matrix calculation for small beamlets is made possible by adopting a scheme that allows concurrent calculation of multiple beamlets of single port. A finite-sized point source (FSPS) beam model is introduced for easy and accurate beam modeling. A DVH based objective function and a parallel platform based algorithm are developed for the optimization of intensities. The calculation accuracy of improved MCNP code and FSPS beam model is validated by dose measurements in phantoms. Agreements better than 1.5% or 0.2 cm have been achieved. Applications of the implemented model to clinical cases of brain, head/neck, lung, spine, pancreas and prostate have demonstrated the feasibility and capability of Monte Carlo based inverse planning for clinical IMRT. Dose distributions of selected treatment plans from a commercial non-Monte Carlo based system are evaluated in comparison with Monte Carlo based calculations. Systematic errors of up to 12% in tumor doses and up to 17% in critical structure doses have been observed. The clinical importance of Monte Carlo based
Asghar Mesbahi
2015-09-01
Full Text Available Introduction Radiotherapy with small fields is used widely in newly developed techniques. Additionally, dose calculation accuracy of treatment planning systems in small fields plays a crucial role in treatment outcome. In the present study, dose calculation accuracy of two commercial treatment planning systems was evaluated against Monte Carlo method. Materials and Methods Siemens Once or linear accelerator was simulated, using MCNPX Monte Carlo code, according to manufacturer’s instructions. Three analytical algorithms for dose calculation including full scatter convolution (FSC in TiGRT, along with convolution and superposition in XiO system were evaluated for a small solid liver tumor. This solid tumor with a diameter of 1.8 cm was evaluated in a thorax phantom, and calculations were performed for different field sizes (1×1, 2×2, 3×3 and4×4 cm2. The results obtained in these treatment planning systems were compared with calculations by MC method (regarded as the most reliable method. Results For FSC and convolution algorithm, comparison with MC calculations indicated dose overestimations of up to 120%and 25% inside the lung and tumor, respectively in 1×1 cm2field size, using an 18 MV photon beam. Regarding superposition, a close agreement was seen with MC simulation in all studied field sizes. Conclusion The obtained results showed that FSC and convolution algorithm significantly overestimated doses of the lung and solid tumor; therefore, significant errors could arise in treatment plans of lung region, thus affecting the treatment outcomes. Therefore, use of MC-based methods and super position is recommended for lung treatments, using small fields and beamlets.
Accuracy assessment of a new Monte Carlo based burnup computer code
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.
Recent developments in the Los Alamos radiation transport code system
Forster, R.A.; Parsons, K. [Los Alamos National Lab., NM (United States)
1997-06-01
A brief progress report on updates to the Los Alamos Radiation Transport Code System (LARTCS) for solving criticality and fixed-source problems is provided. LARTCS integrates the Diffusion Accelerated Neutral Transport (DANT) discrete ordinates codes with the Monte Carlo N-Particle (MCNP) code. The LARCTS code is being developed with a graphical user interface for problem setup and analysis. Progress in the DANT system for criticality applications include a two-dimensional module which can be linked to a mesh-generation code and a faster iteration scheme. Updates to MCNP Version 4A allow statistical checks of calculated Monte Carlo results.
MCT: a Monte Carlo code for time-dependent neutron thermalization problems
In the Monte Carlo simulation of pulse source experiments, the neutron energy spectrum, spatial distribution and total density may be required for a long time after the pulse. If the assemblies are very small, as often occurs in the cases of interest, sophisticated Monte Carlo techniques must be applied which force neutrons to remain in the system during the time interval investigated. In the MCT code a splitting technique has been applied to neutrons exceeding assigned target times, and we have found that this technique compares very favorably with more usual ones, such as the expected leakage probability, giving large gains in computational time and variance. As an example, satisfactory asymptotic thermal spectra with a neutron attenuation of 10-5 were quickly obtained. (U.S.)
MULTI-KENO: a Monte Carlo code for criticality safety analysis
Modifying the Monte Carlo code KENO-IV, the MULTI-KENO code was developed for criticality safety analysis. The following functions were added to the code; (1) to divide a system into many sub-systems named super boxes where the size of box types in each super box can be selected independently, (2) to output graphical view of a system for examining geometrical input data, (3) to solve fixed source problems, (4) to permit intersection of core boundaries and inner geometries, (5) to output ANISN type neutron balance table. With the above function (1), many cases which had to be applied a general geometry option of KENO-IV, became to be treated as box type geometry. In such a case, input data became simpler and required computer time became shorter than those of KENO-IV. This code is now available for the FACOM-M200 computer and the CDC 6600 computer. This report is a computer code manual for MULTI-KENO. (author)
ITS Version 6 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes.
Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William
2008-04-01
ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of lineartime-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 90. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.
Two methods of calculating criticality are available in the 3D generalised geometry Monte Carlo particle transport code SPARTAN (Bending and Heffer, 1975). The first is a matrix technique in which the multiplication constant and source distribution of the system under study are calculated from estimates of fission probabilities and the second a method in which the multiplication constant is inferred from estimates of changes in neutron population over a number of neutron generations. Modifications are described which have been made to the way in which these methods are used in SPARTAN in order to improve the efficiency of criticality calculations. (author)
FORTRAN code-evaluation system
Capps, J. D.; Kleir, R.
1977-01-01
Automated code evaluation system can be used to detect coding errors and unsound coding practices in any ANSI FORTRAN IV source code before they can cause execution-time malfunctions. System concentrates on acceptable FORTRAN code features which are likely to produce undesirable results.
An EGS4 Monte Carlo user code for radiation therapy planning
An EGS4 Monte Carlo user code (the UCRTP code) with voxel geometry has been developed as a prototype of the dose calculation engine for radiation therapy planning. A series of dose calculations for photon beam irradiation to a simplified heterogenous voxel phantom of a lung cancer patient has shown that significant build-up in lung tumor and build-down in surrounding normal lung tissue region exist due to the heterogeneity of the media and small field size. Most of the heterogeneity correction algorithms employed by the current commercial treatment planning systems are not satisfactory enough to account for the build-up/down. Since the commercial systems may significantly underestimate the dose in normal lung tissues, sufficient verification and quality assurance of the radiation therapy planning is needed especially in the lung cancer treatment. (author)
Current status of safety analysis code MARS and uncertainty quantification by Monte-Carlo method
MARS (Multi-dimensional Analysis of Reactor Safety) code has been developed since 1997 for a realistic multi-dimensional thermal-hydraulic system analysis of light water reactor transients. The backbones of MARS are the RELAP5/MOD3.2.1.2 and COBRA-TF codes of USNRC. These two codes were consolidated into a single code by integrating the hydrodynamic solution schemes. New multidimensional TH model has been developed and extended to enable integrated coupled TH analysis through code coupling technique, DLL. The motivation for uncertainty quantification of MARS is considered twofold, 1) to provide “best estimate plus uncertainty” analysis for licensing of commercial power reactor with realistic margins, and 2) to provide support to design and/or validation related analysis for research and production reactors. An assessment of the current LBLOCA uncertainty analysis methodology has been done using data from an integral thermal-hydraulic experiment LOFT L2-5. Monte Carlo calculation has been performed and compared with the tolerance level determined by Wilks formula. The calculation has been done within reasonable CPU time on PC cluster system. Monte-Carlo exercise shows that the 95% upper limit value can be obtained well with 95% confidence level by Wilks formula, although we have to endure 5% risk of PCT under-prediction. The result also shows the statistical fluctuation of limit value using Wilks 1st order is as large as PCT uncertainty itself. The main conclusion is that it is desirable to increase the order of Wilks formula to be higher than the second order to get the reliable safety margin of current design feature. (author)
Coupling CFD code with system code and neutron kinetic code
Vyskocil, Ladislav, E-mail: Ladislav.Vyskocil@ujv.cz; Macek, Jiri
2014-11-15
Highlights: • Coupling interface between CFD code Fluent and system code Athlet was created. • Athlet code is internally coupled with neutron kinetic code Dyn3D. • Explicit coupling of overlapped computational domains was used. • A coupled system of Athlet/Dyn3D+Fluent codes was successfully tested on a real case. - Abstract: The aim of this work was to develop the coupling interface between CFD code Fluent and system code Athlet internally coupled with neutron kinetic code Dyn3D. The coupling interface is intended for simulation of complex transients such as Main Steam Line Break scenarios, which cannot be modeled separately first by system and neutron kinetic code and then by CFD code, because of the feedback between the codes. In the first part of this article, the coupling method is described. Explicit coupling of overlapped computational domains is used in this work. The second part of the article presents a demonstration simulation performed by the coupled system of Athlet/Dyn3D and Fluent. The “Opening a Steam Dump to the Atmosphere” test carried out at the Temelin NPP (VVER-1000) was simulated by the coupled system. In this simulation, the primary and secondary circuits were modeled by Athlet, mixing in downcomer and lower plenum was simulated by Fluent and heat generation in the core was calculated by Dyn3D. The results of the simulation with Athlet/Dyn3D+Fluent were compared with the experimental data and the results from a calculation performed with Athlet/Dyn3D without Fluent.
Coupling CFD code with system code and neutron kinetic code
Highlights: • Coupling interface between CFD code Fluent and system code Athlet was created. • Athlet code is internally coupled with neutron kinetic code Dyn3D. • Explicit coupling of overlapped computational domains was used. • A coupled system of Athlet/Dyn3D+Fluent codes was successfully tested on a real case. - Abstract: The aim of this work was to develop the coupling interface between CFD code Fluent and system code Athlet internally coupled with neutron kinetic code Dyn3D. The coupling interface is intended for simulation of complex transients such as Main Steam Line Break scenarios, which cannot be modeled separately first by system and neutron kinetic code and then by CFD code, because of the feedback between the codes. In the first part of this article, the coupling method is described. Explicit coupling of overlapped computational domains is used in this work. The second part of the article presents a demonstration simulation performed by the coupled system of Athlet/Dyn3D and Fluent. The “Opening a Steam Dump to the Atmosphere” test carried out at the Temelin NPP (VVER-1000) was simulated by the coupled system. In this simulation, the primary and secondary circuits were modeled by Athlet, mixing in downcomer and lower plenum was simulated by Fluent and heat generation in the core was calculated by Dyn3D. The results of the simulation with Athlet/Dyn3D+Fluent were compared with the experimental data and the results from a calculation performed with Athlet/Dyn3D without Fluent
Calibration and simulation of a HPGe well detector using Monte Carlo computer code
Monte Carlo methods are often used in simulating physical and mathematical systems. This computer code is a class of computational algorithms that rely on repeated random sampling to compute their results. Because of their reliance on repeated computation of random or pseudo-random numbers, these methods are most suited to calculation by a computer and tend to be used when it is unfeasible or impossible to compute an exact result with a deterministic algorithm. The Monte Carlo method is used to determine a detector's response curves which are difficult to obtain experimentally. It deals with random numbers for the simulation of the decay conditions and angle of incidence at a given energy value, studying, thus, the random behavior of the radiation, providing response and efficiency curves. The MCNP5 computer code provides means to simulate gamma ray detectors and has been used for this work for the 50keV - 2000 keV energy range. The HPGe well detector was simulated with the MCNP5 computer code and compared with experimental data. The dimensions of both dead layer and the transition layer were determined, and the response curve for a particular geometry was then obtained and compared with the experimental results, in order to verify the detector's simulation. Both results were in very good agreement. (author)
Analysis of the tritium breeding ratio benchmark experiments using the Monte Carlo code TRIPOLI-4
Tritium breeding is an essential element of fusion nuclear technology. A tritium breeding ratio greater than unity is necessary for self-sufficient fueling. To simulate the 14 MeV neutron transport in tritium breeding systems from the D-T fusion reaction, the 3D realistic modeling with Monte Carlo code and the point-wise nuclear data are recommended. Continuous-energy TRIPOLI-4 Monte Carlo transport code has been widely used on the radiation shielding, criticality safety, and fission reactor physics. For supporting the ITER TBM (test blanket module) neutronics study with TRIPOLI-4 code, this paper presents the TRIPOLI-4 simulation of TBR (tritium breeding ratio) for six OKTAVIAN spherical assemblies of Osaka University: Li, Li-C, Pb-Li, Pb-Li-C, Be-Li, and Be-Li-C. It also investigates the impact of nuclear data libraries on TBR calculations from ENDF/B-VI.4, ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, and FENDL-2.1. In general, TRIPOLI-4 produced satisfactory C/E values. Only beryllium of JEFF-3.1 library introduces higher uncertainties.
Recent developments of JAEA’s Monte Carlo code MVP for reactor physics applications
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
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.)
Development of Monte Carlo-based pebble bed reactor fuel management code
Highlights: • A new Monte Carlo-based fuel management code for OTTO cycle pebble bed reactor was developed. • The double-heterogeneity was modeled using statistical method in MVP-BURN code. • The code can perform analysis of equilibrium and non-equilibrium phase. • Code-to-code comparisons for Once-Through-Then-Out case were investigated. • Ability of the code to accommodate the void cavity was confirmed. - Abstract: A fuel management code for pebble bed reactors (PBRs) based on the Monte Carlo method has been developed in this study. The code, named Monte Carlo burnup analysis code for PBR (MCPBR), enables a simulation of the Once-Through-Then-Out (OTTO) cycle of a PBR from the running-in phase to the equilibrium condition. In MCPBR, a burnup calculation based on a continuous-energy Monte Carlo code, MVP-BURN, is coupled with an additional utility code to be able to simulate the OTTO cycle of PBR. MCPBR has several advantages in modeling PBRs, namely its Monte Carlo neutron transport modeling, its capability of explicitly modeling the double heterogeneity of the PBR core, and its ability to model different axial fuel speeds in the PBR core. Analysis at the equilibrium condition of the simplified PBR was used as the validation test of MCPBR. The calculation results of the code were compared with the results of diffusion-based fuel management PBR codes, namely the VSOP and PEBBED codes. Using JENDL-4.0 nuclide library, MCPBR gave a 4.15% and 3.32% lower keff value compared to VSOP and PEBBED, respectively. While using JENDL-3.3, MCPBR gave a 2.22% and 3.11% higher keff value compared to VSOP and PEBBED, respectively. The ability of MCPBR to analyze neutron transport in the top void of the PBR core and its effects was also confirmed
MCNP4B-GN, Monte Carlo Code System for (gamma,n) production and transport in high-Z materials
1 - Description of program or function: MCNP4B-GN is used to treat (gamma,n) production and transport in medical accelerator heads, to study the undesired neutron dose to patients, employing a single code for both the electromagnetic and the neutron transport. 2 - Methods: The code simulates the production of giant dipole resonance (GDR) photoneutrons in thick layers of high-Z elements. Neutrons are generated through evaporation of the compound nucleus or through direct channel; the photoneutron origin coordinates, evaluated as the electromagnetic shower develops are taken into account as well as the energy spectrum of the generated photoneutron. Photoneutron production routines have been inserted into MCNP4B, thus allowing a handling of complicated geometries with a single input definition, a fundamental requirement for this kind of application. The modifications to the standard MCNP4B were made as a 'patch' (i.e. a series of instructions on how to modify the basic code) which is distributed together with a preprocessor. The preprocessor reads the patch, reads MCNP4B and writes MCNP4B-GN. A new subroutine 'GAMMN' has been written, analogous to the subroutine 'ACEGAM' for (n,gamma). GAMMN is called from subroutine 'COLIDP' at a photon collision with one of the 6 elements in question when the energy of the photon is such that there is a non-zero probability of producing a neutron. In analogy with the (n,gamma) capability, the production or not of a neutron is not correlated with the subsequent history of the photon and in particular the choice of the type of photon collision at that spatial point. 3 - Restrictions on the complexity of the problem: The upper energy limit, imposed by theoretical considerations, is 30 MeV for photons and 20 MeV for photoneutrons. The physical model breaks down at higher photon energies, whilst the neutron transport performed with MCNP cannot be simulated for neutron energies greater than 20 MeV (the neutron energy regime in MCNP is
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
Coded aperture coherent scatter imaging for breast cancer detection: a Monte Carlo evaluation
Lakshmanan, Manu N.; Morris, Robert E.; Greenberg, Joel A.; Samei, Ehsan; Kapadia, Anuj J.
2016-03-01
It is known that conventional x-ray imaging provides a maximum contrast between cancerous and healthy fibroglandular breast tissues of 3% based on their linear x-ray attenuation coefficients at 17.5 keV, whereas coherent scatter signal provides a maximum contrast of 19% based on their differential coherent scatter cross sections. Therefore in order to exploit this potential contrast, we seek to evaluate the performance of a coded- aperture coherent scatter imaging system for breast cancer detection and investigate its accuracy using Monte Carlo simulations. In the simulations we modeled our experimental system, which consists of a raster-scanned pencil beam of x-rays, a bismuth-tin coded aperture mask comprised of a repeating slit pattern with 2-mm periodicity, and a linear-array of 128 detector pixels with 6.5-keV energy resolution. The breast tissue that was scanned comprised a 3-cm sample taken from a patient-based XCAT breast phantom containing a tomosynthesis- based realistic simulated lesion. The differential coherent scatter cross section was reconstructed at each pixel in the image using an iterative reconstruction algorithm. Each pixel in the reconstructed image was then classified as being either air or the type of breast tissue with which its normalized reconstructed differential coherent scatter cross section had the highest correlation coefficient. Comparison of the final tissue classification results with the ground truth image showed that the coded aperture imaging technique has a cancerous pixel detection sensitivity (correct identification of cancerous pixels), specificity (correctly ruling out healthy pixels as not being cancer) and accuracy of 92.4%, 91.9% and 92.0%, respectively. Our Monte Carlo evaluation of our experimental coded aperture coherent scatter imaging system shows that it is able to exploit the greater contrast available from coherently scattered x-rays to increase the accuracy of detecting cancerous regions within the breast.
Accumulative Landings System Code Tables
National Oceanic and Atmospheric Administration, Department of Commerce — Code Tables Used In Landings System. These tables assign meanings to the codes that appear in the data tables. Code tables exist for species, gear, state, county,...
Development of a Monte-Carlo Radiative Transfer Code for the Juno/JIRAM Limb Measurements
Sindoni, G.; Adriani, A.; Mayorov, B.; Aoki, S.; Grassi, D.; Moriconi, M.; Oliva, F.
2013-09-01
The Juno/JIRAM instrument will acquire limb spectra of the Jupiter atmosphere in the infrared spectral range. The analysis of these spectra requires a radiative transfer code that takes into account the multiple scattering by particles in a spherical-shell atmosphere. Therefore, we are developing a code based on the Monte-Carlo approach to simulate the JIRAM observations. The validation of the code was performed by comparison with DISORT-based codes.
Application of Monte Carlo code EGS4 to calculate gamma exposure buildup factors
Exposure buildup factors up to 40 mean free paths ranging from 0.015 MeV to 15 MeV photon energy were calculated by using the Monte Carlo simulation code EGS4 for ordinary concrete. The calculation involves PHOTX cross section library, a point isotropic source, infinite uniform medium model and a particle splitting method and considers the Bremsstrahlung, fluorescent effect, correlative (Rayleigh) scatter. The results were compared with the relevant data. Results show that the data of the buildup factors calculated by the Monte Carlo code EGS4 was reliable. The Monte Carlo method can be used widely to calculate gamma-ray exposure buildup factors. (authors)
A 3DHZETRN Code in a Spherical Uniform Sphere with Monte Carlo Verification
Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.
2014-01-01
The computationally efficient HZETRN code has been used in recent trade studies for lunar and Martian exploration and is currently being used in the engineering development of the next generation of space vehicles, habitats, and extra vehicular activity equipment. A new version (3DHZETRN) capable of transporting High charge (Z) and Energy (HZE) and light ions (including neutrons) under space-like boundary conditions with enhanced neutron and light ion propagation is under development. In the present report, new algorithms for light ion and neutron propagation with well-defined convergence criteria in 3D objects is developed and tested against Monte Carlo simulations to verify the solution methodology. The code will be available through the software system, OLTARIS, for shield design and validation and provides a basis for personal computer software capable of space shield analysis and optimization.
Point KENO V.a: A continuous-energy Monte Carlo code for transport applications
KENO V.a is a multigroup Monte Carlo code that solves the Boltzmann transport equation and is used extensively in the criticality safety community to calculate the effective multiplication factor of systems with fissionable material. In this work, a continuous-energy or pointwise version of KENO V.a has been developed by first designing a new continuous-energy cross-section format and then by developing the appropriate Monte Carlo transport procedures to sample the new cross-section format. In order to generate pointwise cross sections for a test library, a series of cross-section processing modules were developed and used to process 50 ENDF/B-VI Release 7 nuclides for the test library. Once the cross-section processing procedures were in place, a continuous-energy version of KENO V.a was developed and tested by calculating 46 test cases that include critical and calculational benchmark problems. The Point KENO-calculated results for the test problems are in agreement with calculated results obtained with the multigroup version of KENO V.a and MCNP4C. Based on the calculated results with the prototypic cross-section library, a continuous-energy version of the KENO V.a code has been successfully developed and demonstrated for modeling systems with fissionable material. (authors)
The Electron-Gamma Shower (EGS) code system is a general purpose package for the Monte Carlo simulation of the coupled transport of electrons and photons in an arbitrary geometry for particles with energies above a few keV up to several hundred GeV (depending on the atomic numbers of the target materials). This report introduces a new, enhanced version called EGS5. In addition to explaining and documenting the various enhancements and changes to the previous version (EGS4), this document includes several introductory and advanced tutorials on the use of EGS5, and also contains the EGS5 User Manual. Our intention has been to make this document wholly self-contained so that the user need not refer to the original EGS4 manual (SLAC-265) in order to use the code. To this end, we have taken the liberty of incorporating into Chapter 2 of this report those portions of Chapter 2 of SLAC-265 which describe physics models of EGS4 retained by EGS5, thereby documenting all the physics contained in EGS5. (author)
The analog linear interpolation approach for Monte Carlo simulation of PGNAA: The CEARPGA code
Zhang, Wenchao; Gardner, Robin P.
2004-01-01
The analog linear interpolation approach (ALI) has been developed and implemented to eliminate the big weight problem in the Monte Carlo simulation code CEARPGA. The CEARPGA code was previously developed to generate elemental library spectra for using the Monte Carlo - library least-squares (MCLLS) approach in prompt gamma-ray neutron activation analysis (PGNAA). In addition, some other improvements to this code have been introduced, including (1) adopting the latest photon cross-section data, (2) using an improved detector response function, (3) adding the neutron activation backgrounds, (4) generating the individual natural background libraries, (5) adding the tracking of annihilation photons from pair production interactions outside of the detector and (6) adopting a general geometry package. The simulated result from the new CEARPGA code is compared with those calculated from the previous CEARPGA code and the MCNP code and experimental data. The new CEARPGA code is found to give the best result.
Running the EGS4 Monte Carlo code with Fortran 90 on a pentium computer
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
On the inner workings of Monte Carlo codes
D. Dubbeldam; A. Torres Knoop; K.S. Walton
2013-01-01
We review state-of-the-art Monte Carlo (MC) techniques for computing fluid coexistence properties (Gibbs simulations) and adsorption simulations in nanoporous materials such as zeolites and metal-organic frameworks. Conventional MC is discussed and compared to advanced techniques such as reactive MC
Data libraries as a collaborative tool across Monte Carlo codes
Augelli, Mauro; Han, Mincheol; Hauf, Steffen; Kim, Chan-Hyeung; Kuster, Markus; Pia, Maria Grazia; Quintieri, Lina; Saracco, Paolo; Seo, Hee; Sudhakar, Manju; Eidenspointner, Georg; Zoglauer, Andreas
2010-01-01
The role of data libraries in Monte Carlo simulation is discussed. A number of data libraries currently in preparation are reviewed; their data are critically examined with respect to the state-of-the-art in the respective fields. Extensive tests with respect to experimental data have been performed for the validation of their content.
Matthew Ellis; Derek Gaston; Benoit Forget; Kord Smith
2011-07-01
In recent years the use of Monte Carlo methods for modeling reactors has become feasible due to the increasing availability of massively parallel computer systems. One of the primary challenges yet to be fully resolved, however, is the efficient and accurate inclusion of multiphysics feedback in Monte Carlo simulations. The research in this paper presents a preliminary coupling of the open source Monte Carlo code OpenMC with the open source Multiphysics Object-Oriented Simulation Environment (MOOSE). The coupling of OpenMC and MOOSE will be used to investigate efficient and accurate numerical methods needed to include multiphysics feedback in Monte Carlo codes. An investigation into the sensitivity of Doppler feedback to fuel temperature approximations using a two dimensional 17x17 PWR fuel assembly is presented in this paper. The results show a functioning multiphysics coupling between OpenMC and MOOSE. The coupling utilizes Functional Expansion Tallies to accurately and efficiently transfer pin power distributions tallied in OpenMC to unstructured finite element meshes used in MOOSE. The two dimensional PWR fuel assembly case also demonstrates that for a simplified model the pin-by-pin doppler feedback can be adequately replicated by scaling a representative pin based on pin relative powers.
Aurora T: a Monte Carlo code for transportation of neutral atoms in a toroidal plasma
This paper contains a short description of Aurora code. This code have been developed at Princeton with Monte Carlo method for calculating neutral gas in cylindrical plasma. In this work subroutines such one can take in account toroidal geometry are developed
Gao, Wen
2015-01-01
This comprehensive and accessible text/reference presents an overview of the state of the art in video coding technology. Specifically, the book introduces the tools of the AVS2 standard, describing how AVS2 can help to achieve a significant improvement in coding efficiency for future video networks and applications by incorporating smarter coding tools such as scene video coding. Topics and features: introduces the basic concepts in video coding, and presents a short history of video coding technology and standards; reviews the coding framework, main coding tools, and syntax structure of AV
Nelson, Adam
Multi-group scattering moment matrices are critical to the solution of the multi-group form of the neutron transport equation, as they are responsible for describing the change in direction and energy of neutrons. These matrices, however, are difficult to correctly calculate from the measured nuclear data with both deterministic and stochastic methods. Calculating these parameters when using deterministic methods requires a set of assumptions which do not hold true in all conditions. These quantities can be calculated accurately with stochastic methods, however doing so is computationally expensive due to the poor efficiency of tallying scattering moment matrices. This work presents an improved method of obtaining multi-group scattering moment matrices from a Monte Carlo neutron transport code. This improved method of tallying the scattering moment matrices is based on recognizing that all of the outgoing particle information is known a priori and can be taken advantage of to increase the tallying efficiency (therefore reducing the uncertainty) of the stochastically integrated tallies. In this scheme, the complete outgoing probability distribution is tallied, supplying every one of the scattering moment matrices elements with its share of data. In addition to reducing the uncertainty, this method allows for the use of a track-length estimation process potentially offering even further improvement to the tallying efficiency. Unfortunately, to produce the needed distributions, the probability functions themselves must undergo an integration over the outgoing energy and scattering angle dimensions. This integration is too costly to perform during the Monte Carlo simulation itself and therefore must be performed in advance by way of a pre-processing code. The new method increases the information obtained from tally events and therefore has a significantly higher efficiency than the currently used techniques. The improved method has been implemented in a code system
Development of a space radiation Monte Carlo computer simulation based on the FLUKA and ROOT codes.
Pinsky, L S; Wilson, T L; Ferrari, A; Sala, P; Carminati, F; Brun, R
2001-01-01
This NASA funded project is proceeding to develop a Monte Carlo-based computer simulation of the radiation environment in space. With actual funding only initially in place at the end of May 2000, the study is still in the early stage of development. The general tasks have been identified and personnel have been selected. The code to be assembled will be based upon two major existing software packages. The radiation transport simulation will be accomplished by updating the FLUKA Monte Carlo program, and the user interface will employ the ROOT software being developed at CERN. The end-product will be a Monte Carlo-based code which will complement the existing analytic codes such as BRYNTRN/HZETRN presently used by NASA to evaluate the effects of radiation shielding in space. The planned code will possess the ability to evaluate the radiation environment for spacecraft and habitats in Earth orbit, in interplanetary space, on the lunar surface, or on a planetary surface such as Mars. Furthermore, it will be useful in the design and analysis of experiments such as ACCESS (Advanced Cosmic-ray Composition Experiment for Space Station), which is an Office of Space Science payload currently under evaluation for deployment on the International Space Station (ISS). FLUKA will be significantly improved and tailored for use in simulating space radiation in four ways. First, the additional physics not presently within the code that is necessary to simulate the problems of interest, namely the heavy ion inelastic processes, will be incorporated. Second, the internal geometry package will be replaced with one that will substantially increase the calculation speed as well as simplify the data input task. Third, default incident flux packages that include all of the different space radiation sources of interest will be included. Finally, the user interface and internal data structure will be melded together with ROOT, the object-oriented data analysis infrastructure system. Beyond
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.)
OpenMC: A state-of-the-art Monte Carlo code for research and development
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
Code system to compute radiation dose in human phantoms
Monte Carlo photon transport code and a code using Monte Carlo integration of a point kernel have been revised to incorporate human phantom models for an adult female, juveniles of various ages, and a pregnant female at the end of the first trimester of pregnancy, in addition to the adult male used earlier. An analysis code has been developed for deriving recommended values of specific absorbed fractions of photon energy. The computer code system and calculational method are described, emphasizing recent improvements in methods
Monte Carlo N Particle code - Dose distribution of clinical electron beams in inhomogeneous phantoms
H A Nedaie
2013-01-01
Full Text Available Electron dose distributions calculated using the currently available analytical methods can be associated with large uncertainties. The Monte Carlo method is the most accurate method for dose calculation in electron beams. Most of the clinical electron beam simulation studies have been performed using non- MCNP [Monte Carlo N Particle] codes. Given the differences between Monte Carlo codes, this work aims to evaluate the accuracy of MCNP4C-simulated electron dose distributions in a homogenous phantom and around inhomogeneities. Different types of phantoms ranging in complexity were used; namely, a homogeneous water phantom and phantoms made of polymethyl methacrylate slabs containing different-sized, low- and high-density inserts of heterogeneous materials. Electron beams with 8 and 15 MeV nominal energy generated by an Elekta Synergy linear accelerator were investigated. Measurements were performed for a 10 cm × 10 cm applicator at a source-to-surface distance of 100 cm. Individual parts of the beam-defining system were introduced into the simulation one at a time in order to show their effect on depth doses. In contrast to the first scattering foil, the secondary scattering foil, X and Y jaws and applicator provide up to 5% of the dose. A 2%/2 mm agreement between MCNP and measurements was found in the homogenous phantom, and in the presence of heterogeneities in the range of 1-3%, being generally within 2% of the measurements for both energies in a "complex" phantom. A full-component simulation is necessary in order to obtain a realistic model of the beam. The MCNP4C results agree well with the measured electron dose distributions.
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
Using deterministic codes to accelerate continuous energy Monte-Carlo standards calculations
Deterministic codes are usually used for critical parameters or one dimension geometry calculations. Advantages of the use of deterministic codes are speed of the calculation and the absence of standard deviation on the keff results. Nevertheless, the deterministic results are affected by several intrinsic uncertainties as energetic condensation or self-shielding. So the way to proceed at CEA expert criticality group (CEA/SERMA/CP2C) is to always check the main results (minimum critical or maximal permissible values and un-moderated values) with a punctual Monte Carlo calculation. These last years, in particular cases (pure actinide fissile media, exotic reflectors), large discrepancies have been observed between the keff calculated by the CRISTAL V1 route reference (continuous energy Monte Carlo code TRIPOLI-4) and the keff target (by the standard route APOLLO2-Sn). The problematic for these cases was how to transpose the keff discrepancies observed between standard and reference routes to the dimensions (mass, thickness...) or how to reduce the keff discrepancies using optimized options of the deterministic code. One solution to transpose discrepancies is to iterate on dimensions using a punctual Monte Carlo code to achieve the desired keff eigenvalue. But, the amount of time for obtaining a good standard deviation and also the desired keff eigenvalue inside the Monte Carlo calculation uncertainty can quickly increase. The principle of the method presented in this paper is that the discrepancy between deterministic code and Monte-Carlo code, calculated at the same dimension, is low variable with the dimension. Therefore, correcting the keff eigenvalue on which the deterministic code converge with the discrepancy observed, leads to a dimension nearer to the true dimension (i.e. the dimension where Monte-Carlo code keff calculation is close to the keff eigenvalue). If the keff eigenvalue is outside the Monte Carlo uncertainty, the discrepancy is recalculated and
Calculations of neutron penetration through graphite medium with Monte Carlo code MCNP
Experiments for fast neutron penetration through graphite are analysed with the continuous energy Monte Carlo code MCNP. Reaction rates and energy spectra obtained with the MCNP are compared with measured values and calculated ones with McBEND code. And validity of penetration calculation with the MCNP is comfirmed. In addition, it is revealed that the MCNP code using Weight-Window method is well applicable to calculations of neutron penetration through graphite up to 70 cm in depth. (author)
Monte-Carlo Simulation for PDC-Based Optical CDMA System
FAHIM AZIZ UMRANI; AHSAN AHMED URSANI; ABDUL WAHEED UMRANI
2010-01-01
This paper presents the Monte-Carlo simulation of Optical CDMA (Code Division Multiple Access) systems, and analyse its performance in terms of the BER (Bit Error Rate). The spreading sequence chosen for CDMA is Perfect Difference Codes. Furthermore, this paper derives the expressions of noise variances from first principles to calibrate the noise for both bipolar (electrical domain) and unipolar (optical domain) signalling required for Monte-Carlo simulation. The simulated res...
Highlights: • We present a new Monte Carlo method to perform sensitivity/perturbation calculations. • Sensitivity of keff, reaction rates, point kinetics parameters to nuclear data. • Fully continuous implicitly constrained Monte Carlo sensitivities to scattering distributions. • Implementation of the method in the continuous energy Monte Carlo code SERPENT. • Verification against ERANOS and TSUNAMI generalized perturbation theory results. - Abstract: In this work, the implementation of a collision history-based approach to sensitivity/perturbation calculations in the Monte Carlo code SERPENT is discussed. The proposed methods allow the calculation of the effects of nuclear data perturbation on several response functions: the effective multiplication factor, reaction rate ratios and bilinear ratios (e.g., effective kinetics parameters). SERPENT results are compared to ERANOS and TSUNAMI Generalized Perturbation Theory calculations for two fast metallic systems and for a PWR pin-cell benchmark. New methods for the calculation of sensitivities to angular scattering distributions are also presented, which adopts fully continuous (in energy and angle) Monte Carlo estimators
MCNP: a general Monte Carlo code for neutron and photon transport
The general-purpose Monte Carlo code MCNP ca be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation are accounted for. Thermal neutrons are described by both the free-gas and S(α,β) models. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. MCNP includes an elaborate, interactive plotting capability that allows the user to view his input geometry to help check for setup errors. Standard features which are available to improve computational efficiency include geometry splitting and Russian roulette, weight cutoff with Russian roulette, correlated sampling, analog capture or capture by weight reduction, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point or ring detectors, deterministically transporting pseudo-particles to designated regions, track-length estimators, source biasing, and several parameter cutoffs. Extensive summary information is provided to help the user better understand the physics and Monte Carlo simulation of his problem. The standard, user-defined output of MCNP includes two-way current as a function of direction across any set of surfaces or surface segments in the problem. Flux across any set of surfaces or surface segments is available. 58 figures, 28 tables
SU-E-T-578: MCEBRT, A Monte Carlo Code for External Beam Treatment Plan Verifications
Chibani, O; Ma, C [Fox Chase Cancer Center, Philadelphia, PA (United States); Eldib, A [Fox Chase Cancer Center, Philadelphia, PA (United States); Al-Azhar University, Cairo (Egypt)
2014-06-01
Purpose: Present a new Monte Carlo code (MCEBRT) for patient-specific dose calculations in external beam radiotherapy. The code MLC model is benchmarked and real patient plans are re-calculated using MCEBRT and compared with commercial TPS. Methods: MCEBRT is based on the GEPTS system (Med. Phys. 29 (2002) 835–846). Phase space data generated for Varian linac photon beams (6 – 15 MV) are used as source term. MCEBRT uses a realistic MLC model (tongue and groove, rounded ends). Patient CT and DICOM RT files are used to generate a 3D patient phantom and simulate the treatment configuration (gantry, collimator and couch angles; jaw positions; MLC sequences; MUs). MCEBRT dose distributions and DVHs are compared with those from TPS in absolute way (Gy). Results: Calculations based on the developed MLC model closely matches transmission measurements (pin-point ionization chamber at selected positions and film for lateral dose profile). See Fig.1. Dose calculations for two clinical cases (whole brain irradiation with opposed beams and lung case with eight fields) are carried out and outcomes are compared with the Eclipse AAA algorithm. Good agreement is observed for the brain case (Figs 2-3) except at the surface where MCEBRT dose can be higher by 20%. This is due to better modeling of electron contamination by MCEBRT. For the lung case an overall good agreement (91% gamma index passing rate with 3%/3mm DTA criterion) is observed (Fig.4) but dose in lung can be over-estimated by up to 10% by AAA (Fig.5). CTV and PTV DVHs from TPS and MCEBRT are nevertheless close (Fig.6). Conclusion: A new Monte Carlo code is developed for plan verification. Contrary to phantombased QA measurements, MCEBRT simulate the exact patient geometry and tissue composition. MCEBRT can be used as extra verification layer for plans where surface dose and tissue heterogeneity are an issue.
MCMG: a 3-D multigroup P3 Monte Carlo code and its benchmarks
In this paper a 3-D Monte Carlo multigroup neutron transport code MCMG has been developed from a coupled neutron and photon transport Monte Carlo code MCNP. The continuous-energy cross section library of the MCNP code is replaced by the multigroup cross section data generated by the transport lattice code, such as the WIMS code. It maintains the strong abilities of MCNP for geometry treatment, counting, variance reduction techniques and plotting. The multigroup neutron scattering cross sections adopt the Pn (n ≤ 3) approximation. The test results are in good agreement with the results of other methods and experiments. The number of energy groups can be varied from few groups to multigroup, and either macroscopic or microscopic cross section can be used. (author)
A vectorized Monte Carlo code for modeling photon transport in SPECT
A vectorized Monte Carlo computer code has been developed for modeling photon transport in single photon emission computed tomography (SPECT). The code models photon transport in a uniform attenuating region and photon detection by a gamma camera. It is adapted from a history-based Monte Carlo code in which photon history data are stored in scalar variables and photon histories are computed sequentially. The vectorized code is written in FORTRAN77 and uses an event-based algorithm in which photon history data are stored in arrays and photon history computations are performed within DO loops. The indices of the DO loops range over the number of photon histories, and these loops may take advantage of the vector processing unit of our Stellar GS1000 computer for pipelined computations. Without the use of the vector processor the event-based code is faster than the history-based code because of numerical optimization performed during conversion to the event-based algorithm. When only the detection of unscattered photons is modeled, the event-based code executes 5.1 times faster with the use of the vector processor than without; when the detection of scattered and unscattered photons is modeled the speed increase is a factor of 2.9. Vectorization is a valuable way to increase the performance of Monte Carlo code for modeling photon transport in SPECT
User's manual of the MKENO-DAR code system
The computer code manual of MKENO-DAR which is a direct angular representation Monte Carlo code for criticality safety analysis is already issued as JAERI-M report, however, complex pre-stage data handlings and calculations by auxiliary programs are required before the execution of MKENO-DAR. The MKENO-DAR CODE SYSTEM widely spans a whole code system including MKENO-DAR and other pre-stage auxiliary programs. This report discusses the systematic treatment of the MKENO-DAR CODE SYSTEM and shows the simplified calculation technique from the user side of view. (author)
Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes
Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code)
贾清刚; 张天奎; 张凤娜; 胡华四
2013-01-01
开发了基于Geant4的Z箍缩中子编码成像系统模拟平台,实现聚变中子编码成像诊断系统各关键部件的完整模拟.获得了低中子产额(约1010量级)下,中子经编码孔编码后在闪烁体阵列中形成的发光分布图像.利用维纳滤波、Richardson-Lucy(RL)及遗传算法(GA)对低中子产额下获得的极低信噪比图像进行重建,并对信噪比、中子产额及重建效果进行了对比研究,结果表明:遗传算法对低信噪比中子编码图像的重建具有很强的鲁棒性;中子编码图像的信噪比与遗传算法重建结果的准确性呈正比.%The model of Z-pinch driven fusion imaging diagnosis system was set up by a Monte Carlo code based on the Geant4 simulation toolkit. All physical processes that the reality involves are taken into consideration in simulation. The light image of low neutron yield (about 1010) pill was obtained. Three types of image reconstruction algorithm, i. e. Richardson-Lucy, Wiener filtering and genetic algorithm were employed to reconstruct the neutron image with a low signal to noise ratio (SNR) and yield. The effects of neutron yields and the SNR on reconstruction performance were discussed. The results show that genetic algorithm is very robust for reconstructing neutron images with a low SNR. And the index of reconstruction performance and the image correlation coefficient using genetic algorithm, are proportional to the SNR of the neutron coded image.
Tokyo Metropolitan University of Health Sciences has done The Information Education using EGS4 Monte Carlo code since the 1998 fiscal year. Two items under practical training item were done. 1. The interaction between photon of 0.1 ∼ 10 MeV (Mega Electron Volt: MeV) and Aluminum (Al), Iron (Fe) and Lead (Pb). 2. The simulation of gamma ray energy measurement of the radiation detector. As the result, the student was possible the understanding of the radiation physics for the easiness at Practical training of EGS4 Monte Carlo code. (author)
The Monte Carlo code MCBEND - where it is and where it's going
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)
Parallelization of MCATNP MONTE CARLO particle transport code by using MPI
A Monte Carlo code for simulating Atmospheric Transport of Neutrons and Photons (MCATNP) is used to simulate the ionization effects caused by high altitude nuclear detonation (HAND) and it was parallelized in MPI by adopting the leap random number producer and modifying the original serial code. The parallel results and serial results are identical. The speedup increases almost linearly with the number of processors used. The parallel efficiency is up to to 97% while 16 processors are used, and 94% while 32 are used. The experimental results show that parallelization can obviously reduce the calculation time of Monte Carlo simulation of HAND ionization effects. (authors)
Generalized Albedo option on the Morse Monte Carlo code
The advisability of using the albedo procedure for solving deep penetration shielding problems which have ducts and other penetrations is investigated. It is generally accepted that the use of albedo data can dramatically improve the computational efficiency of certain Monte Carlo calculations - however the accuracy of these results may be unacceptable because of lost information during the albedo event and serious errors in the available differential albedo data. This study has been done to evaluate and appropriately modify the MORSE/BREESE package, to develop new methods for generating the required albedo data, and to extend the adjoint capability to the albedo modified calculations. The major modifications include the tracking of special particles inside albedo media, an option to displace the point-of-emergence during an albedo event, and an option to read, process, and use spatially-dependent albedo data for both forward and adjoint calculations. (author)
The three-dimensional Monte-Carlo code TRIPOLI-02
TRIPOLI-2 solves the transport equation for neutrons or gamma rays in tridimensional geometrical configurations. TRIPOLI uses the Monte Carlo method. This method allows to treat exactly the geometrical configurations, the energy losses and the scattering laws. TRIPOLI 2 allows to treat the following problems: gamma transport problems, neutrons transport problems with fixed source (the problems can be time dependent or not), critical problems without fixed source and research of multiplication factor due to fissions, subcritical problems with fixed source and with multiplication by fission. These problems can be separate in two types. First type: shielding problems essentially with deep penetration and streaming through voids. Biasing technics are used to reduce the computing time. Second type: core problems for cell calculations or for small core calculations. In this case, it is necessary to have a fine representation of the cross sections. The thermalization is also treated exactly
Importance function by collision probabilities for Monte Carlo code Tripoli
We present a completely automatic biasing technique where the parameters of the biased simulation are deduced from the solution of the adjoint transport equation calculated by collision probabilities. In this study we shall estimate the importance function through collision probabilities method and we shall evaluate its possibilities thanks to a Monte Carlo calculation. We have run simulations with this new biasing method for one-group transport problems with isotropic shocks (one dimension geometry and X-Y geometry) and for multigroup problems with anisotropic shocks (one dimension geometry). For the anisotropic problems we solve the adjoint equation with anisotropic collision probabilities. The results show that for the one-group and homogeneous geometry transport problems the method is quite optimal without Splitting and Russian Roulette technique but for the multigroup and heterogeneous X-Y geometry ones the figures of merit are higher if we add Splitting and Russian Roulette technique
OpenMC: a state-of-the-Art Monte Carlo code for research and development
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)
The FLUKA code for application of Monte Carlo methods to promote high precision ion beam therapy
Parodi, K; Cerutti, F; Ferrari, A; Mairani, A; Paganetti, H; Sommerer, F
2010-01-01
Monte Carlo (MC) methods are increasingly being utilized to support several aspects of commissioning and clinical operation of ion beam therapy facilities. In this contribution two emerging areas of MC applications are outlined. The value of MC modeling to promote accurate treatment planning is addressed via examples of application of the FLUKA code to proton and carbon ion therapy at the Heidelberg Ion Beam Therapy Center in Heidelberg, Germany, and at the Proton Therapy Center of Massachusetts General Hospital (MGH) Boston, USA. These include generation of basic data for input into the treatment planning system (TPS) and validation of the TPS analytical pencil-beam dose computations. Moreover, we review the implementation of PET/CT (Positron-Emission-Tomography / Computed- Tomography) imaging for in-vivo verification of proton therapy at MGH. Here, MC is used to calculate irradiation-induced positron-emitter production in tissue for comparison with the +-activity measurement in order to infer indirect infor...
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
Tripoli-3: monte Carlo transport code for neutral particles - version 3.5 - users manual
The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)
Depletion of a BWR lattice using the racer continuous energy Monte Carlo code
In the past several years there has been a renewed interest in the accuracy of a new generation of lattice physics codes. Most of the time these codes are benchmarked against Monte Carlo codes only at beginning of cycle. In this paper a highly heterogeneous BWR lattice depletion benchmark problem is presented. Results of a 40% void depletion using the RACER continuous energy Monte Carlo code are also presented. Complete problem specifications are given so that comparisons with lattice physics codes or other Monte Carlo codes is possible. The RACER calculations were performed with the ENDF/B-V cross section set. Each flux calculation utilized 2.7 million histories resulting in 95% confidence intervals of ∼1 milli-k on the eigenvalue and ∼1% uncertainties on pin-wise power fractions. Timing statistics for the calculation using the vectorized RACER code averaged ∼ 24,000 neutrons/minute on a single processor of a CRAY-C90 computer
MKENO-DAR: a direct angular representation Monte Carlo code for criticality safety analysis
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)
Recently, sensitivity and uncertainty (S/U) techniques have been used to determine the area of applicability (AOA) of critical experiments used for code and data validation. These techniques require the computation of energy-dependent sensitivity coefficients for multiple reaction types for every nuclide in each system included in the validation. The sensitivity coefficients, as used for this application, predict the relative change in the system multiplication factor due to a relative change in a given cross-section data component or material number density. Thus, a sensitivity coefficient, S, for some macroscopic cross section, Σ, is expressed as S = Σ/k ∂k/∂Σ, where k is the effective neutron multiplication factor for the system. The sensitivity coefficient for the density of a material is equivalent to that of the total macroscopic cross section. Two distinct techniques have been employed in Monte Carlo radiation transport codes for the computation of sensitivity coefficients. The first, and most commonly employed, is the differential sampling technique. The second is the adjoint-based perturbation theory approach. This paper briefly describes each technique and presents the results of a simple test case, pointing out discrepancies in the computed results and proposing a remedy to these discrepancies
A Monte Carlo study of the effect of coded-aperture material and thickness on neutron imaging
In this paper, a coded-aperture design for a scintillator-based neutron imaging system has been simulated using a series of Monte Carlo simulations. Using Monte Carlo simulations, work to optimise a system making use of the EJ-426 neutron scintillator detector has been conducted. This type of scintillator has a low sensitivity to gamma rays and is therefore particularly useful for neutron detection in a mixed radiation environment. Simulations have been conducted using varying coded-aperture materials and different coded-aperture thicknesses. From this, neutron images have been produced, compared qualitatively and quantitatively for each case to find the best material for the MURA (modified uniformly redundant array) pattern. The neutron images generated also allow observations on how differing thicknesses of coded-aperture impact the system. A system in which a neutron sensitive scintillator has been used in conjunction with a MURA coded aperture to detect and locate a neutron emitting point source centralised in the system has been simulated. A comparison between the results of the different coded-aperture thicknesses is discussed, via the calculation of system error between the reconstructed source location and the actual location. As the system is small scale with a relatively large step along the axis (0.5 cm), it is justifiable to say that the smaller error values provide satisfactory results, which correlate with only a few centimetres difference in the reconstructed source location to actual source location. A general trend of increasing error can be deduced both as the thickness of the coded-aperture material decreases and the capture cross section of the different materials reduces. (authors)
Parallel processing of Monte Carlo code MCNP for particle transport problem
Higuchi, Kenji; Kawasaki, Takuji
1996-06-01
It is possible to vectorize or parallelize Monte Carlo codes (MC code) for photon and neutron transport problem, making use of independency of the calculation for each particle. Applicability of existing MC code to parallel processing is mentioned. As for parallel computer, we have used both vector-parallel processor and scalar-parallel processor in performance evaluation. We have made (i) vector-parallel processing of MCNP code on Monte Carlo machine Monte-4 with four vector processors, (ii) parallel processing on Paragon XP/S with 256 processors. In this report we describe the methodology and results for parallel processing on two types of parallel or distributed memory computers. In addition, we mention the evaluation of parallel programming environments for parallel computers used in the present work as a part of the work developing STA (Seamless Thinking Aid) Basic Software. (author)
Development of 3d reactor burnup code based on Monte Carlo method and exponential Euler method
Burnup analysis plays a key role in fuel breeding, transmutation and post-processing in nuclear reactor. Burnup codes based on one-dimensional and two-dimensional transport method have difficulties in meeting the accuracy requirements. A three-dimensional burnup analysis code based on Monte Carlo method and Exponential Euler method has been developed. The coupling code combines advantage of Monte Carlo method in complex geometry neutron transport calculation and FISPACT in fast and precise inventory calculation, meanwhile resonance Self-shielding effect in inventory calculation can also be considered. The IAEA benchmark text problem has been adopted for code validation. Good agreements were shown in the comparison with other participants' results. (authors)
Perfetti, Christopher M [ORNL; Martin, William R [University of Michigan; Rearden, Bradley T [ORNL; Williams, Mark L [ORNL
2012-01-01
Three methods for calculating continuous-energy eigenvalue sensitivity coefficients were developed and implemented into the SHIFT Monte Carlo code within the Scale code package. The methods were used for several simple test problems and were evaluated in terms of speed, accuracy, efficiency, and memory requirements. A promising new method for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was developed and produced accurate sensitivity coefficients with figures of merit that were several orders of magnitude larger than those from existing methods.
Its version 3.0. The integrated TIGER series of coupled electron/photon Monte Carlo transport codes
The ITS system is described, which is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields. It combines operational simplicity and physical accuracy in order to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Flexibility of construction permits tailoring of the codes to specific applications and extension of code capabilities to more complex applications through simple update procedures. (author)
QCDMPI - pure QCD Monte Carlo simulation code with MPI
QCDMPI is a pure QCD simulation code with MPI calls. QCDMPI is very portable because; - you can simulate any-dimensional QCD, - on any-dimensional partitioning, - on any number of processors, - with rather small working area. Also by this program, you can get two performances, - calculation (link update time) - communication (MB/sec). In this paper, outline of QCDMPI is reported. Comparison of the performances on several parallel machines; AP1000, AP1000+, AP3000, Cenju-3, Paragon, SR2201 and Workstation Cluster, is also reported. (orig.)
Verification of the shift Monte Carlo code with the C5G7 reactor benchmark
Sly, N. C.; Mervin, B. T. [Dept. of Nuclear Engineering, Univ. of Tennessee, 311 Pasqua Engineering Building, Knoxville, TN 37996-2300 (United States); Mosher, S. W.; Evans, T. M.; Wagner, J. C. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, 311 Pasqua Engineering Building, Knoxville, TN 37996-2300 (United States)
2012-07-01
Shift is a new hybrid Monte Carlo/deterministic radiation transport code being developed at Oak Ridge National Laboratory. At its current stage of development, Shift includes a parallel Monte Carlo capability for simulating eigenvalue and fixed-source multigroup transport problems. This paper focuses on recent efforts to verify Shift's Monte Carlo component using the two-dimensional and three-dimensional C5G7 NEA benchmark problems. Comparisons were made between the benchmark eigenvalues and those output by the Shift code. In addition, mesh-based scalar flux tally results generated by Shift were compared to those obtained using MCNP5 on an identical model and tally grid. The Shift-generated eigenvalues were within three standard deviations of the benchmark and MCNP5-1.60 values in all cases. The flux tallies generated by Shift were found to be in very good agreement with those from MCNP. (authors)
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)
The present of shielding analysis with nuclear data for continuous energy Monte Carlo code MCNP
Following three problems are analyzed by continuous energy Monte Carlo code MCNP with JENDL-3.2, 3.3, and ENDF/B-VI. 1. Shielding analysis of WINFRITH-Aspins iron deep penetration experiment. 2. Shielding analysis of TN-12A spent fuel transport cask experiment. 3. Shielding analysis of modular shielding house keeping spent fuel transportable casks. (author)
Use of the FLUKA Monte Carlo code for Hadron Therapy Application
Monte Carlo (M C) codes are increasingly spreading in the hadron therapy community due to their detailed description of radiation transport and interaction with matter. M C methods are being utilized at several institutions for a wide range of activities spanning from beam characterization to quality assurance and dosimetric/radiobiological studies. The suitability of a M C code for application to hadron therapy demands accurate and reliable physical models for the description of the transport and the interaction of all components of the expected radiation field (ions, hadrons, electrons, positrons and photons). This becomes extremely important for correctly performing not only physical but also biologically based dose calculations especially in cases where ions heavier than protons are involved. In addition, accurate prediction of emerging secondary radiation is of utmost importance in emerging areas of research aiming to in vivo treatment verification. This contribution will address the specific case of the general-purpose particle and interaction code FLUKA. Validations and applications at several experimental sites as well as proton/ion therapy facilities with active beam delivery systems will be presented: Generation of synchrotron accelerator library of proton/carbon ion beam energies and foci (i.e., lateral widths at the iso centre of the treatment unit). Physical database generation: laterally integrated depth-dose profiles, lateral-dose distributions at different depths, secondary fragments yields and fragment energy spectra at different depths. Forward M C re-calculations of physical/RBE-weighted dose distributions of proton and carbon ion treatment plans. M C-based treatment planning in proton therapy. The satisfactorily agreement of FLUKA against several dosimetric/nuclear yields data indicates that the code already represents a valuable choice for supporting a large variety of applications in proton and ion beam therapy
Evaluation of Monte Carlo Codes Regarding the Calculated Detector Response Function in NDP Method
The basis of the NDP is the irradiation of a sample with a thermal or cold neutron beam and the subsequent release of charged particles due to neutron-induced exoergic charged particle reactions. Neutrons interact with the nuclei of elements and release mono-energetic charged particles, e.g. alpha particles or protons, and recoil atoms. Depth profile of the analyzed element can be obtained by making a linear transformation of the measured energy spectrum by using the stopping power of the sample material. A few micrometer of the material can be analyzed nondestructively, and on the order of 10nm depth resolution can be obtained depending on the material type with NDP method. In the NDP method, the one first steps of the analytical process is a channel-energy calibration. This calibration is normally made with the experimental measurement of NIST Standard Reference Material sample (SRM-93a). In this study, some Monte Carlo (MC) codes were tried to calculate the Si detector response function when this detector accounted the energy charges particles emitting from an analytical sample. In addition, these MC codes were also tried to calculate the depth distributions of some light elements (10B, 3He, 6Li, etc.) in SRM-93a and SRM-2137 samples. These calculated profiles were compared with the experimental profiles and SIMS profiles. In this study, some popular MC neutron transport codes are tried and tested to calculate the detector response function in the NDP method. The simulations were modeled based on the real CN-NDP system which is a part of Cold Neutron Activation Station (CONAS) at HANARO (KAERI). The MC simulations are very successful at predicting the alpha peaks in the measured energy spectrum. The net area difference between the measured and predicted alpha peaks are less than 1%. A possible explanation might be bad cross section data set usage in the MC codes for the transport of low energetic lithium atoms inside the silicon substrate
Subroutines to Simulate Fission Neutrons for Monte Carlo Transport Codes
Lestone, J P
2014-01-01
Fortran subroutines have been written to simulate the production of fission neutrons from the spontaneous fission of 252Cf and 240Pu, and from the thermal neutron induced fission of 239Pu and 235U. The names of these four subroutines are getnv252, getnv240, getnv239, and getnv235, respectively. These subroutines reproduce measured first, second, and third moments of the neutron multiplicity distributions, measured neutron-fission correlation data for the spontaneous fission of 252Cf, and measured neutron-neutron correlation data for both the spontaneous fission of 252Cf and the thermal neutron induced fission of 235U. The codes presented here can be used to study the possible uses of neutron-neutron correlations in the area of transparency measurements and the uses of neutron-neutron correlations in coincidence neutron imaging.
Homma, Yuto; Moriwaki, Hiroyuki; Ohki, Shigeo; Ikeda, Kazumi
2014-06-01
This paper deals with verification of three dimensional triangular prismatic discrete ordinates transport calculation code ENSEMBLE-TRIZ by comparison with multi-group Monte Carlo calculation code GMVP in a large fast breeder reactor. The reactor is a 750 MWe electric power sodium cooled reactor. Nuclear characteristics are calculated at beginning of cycle of an initial core and at beginning and end of cycle of equilibrium core. According to the calculations, the differences between the two methodologies are smaller than 0.0002 Δk in the multi-plication factor, relatively about 1% in the control rod reactivity, and 1% in the sodium void reactivity.
Verification of SMART Neutronics Design Methodology by the MCNAP Monte Carlo Code
SMART is a small advanced integral pressurized water reactor (PWR) of 330 MW(thermal) designed for both electricity generation and seawater desalinization. The CASMO-3/MASTER nuclear analysis system, a design-basis of Korean PWR plants, has been employed for the SMART core nuclear design and analysis because the fuel assembly (FA) characteristics and reactor operating conditions in temperature and pressure are similar to those of PWR plants. However, the SMART FAs are highly poisoned with more than 20 Al2O3-B4C plus additional Gd2O3/UO2 BPRs each FA. The reactor is operated with control rods inserted. Therefore, the flux and power distribution may become more distorted than those of commercial PWR plants. In addition, SMART should produce power from room temperature to hot-power operating condition because it employs nuclear heating from room temperature. This demands reliable predictions of core criticality, shutdown margin, control rod worth, power distributions, and reactivity coefficients at both room temperature and hot operating condition, yet no such data are available to verify the CASMO-3/MASTER (hereafter MASTER) code system. In the absence of experimental verification data for the SMART neutronics design, the Monte Carlo depletion analysis program MCNAP is adopted as near-term alternatives for qualifying MASTER neutronics design calculations. The MCNAP is a personal computer-based continuous energy Monte Carlo neutronics analysis program written in C++ language. We established its qualification by presenting its prediction accuracy on measurements of Venus critical facilities and core neutronics analysis of a PWR plant in operation, and depletion characteristics of integral burnable absorber FAs of the current PWR. Here, we present a comparison of MASTER and MCNAP neutronics design calculations for SMART and establish the qualification of the MASTER system
SFR whole core burnup calculations with TRIPOLI-4 Monte Carlo code
Under the Working Party on Scientific Issues of Reactor Systems (WPRS) of the OECD/NEA, an international collaboration benchmark was recently established on the neutronic analysis of four Sodium-cooled Fast Reactor (SFR) concepts of the Generation- IV nuclear energy systems. As the whole core Monte Carlo depletion calculation is one of the essential challenges of current reactor physics studies, the continuous-energy TRIPOLI-4 Monte Carlo transport code was firstly used in this study to perform whole core 3D neutronic calculations for these four SFR cores. Two medium size (1000 MWt) and two large size (3600 MWt) SFR of GEN-IV systems were analyzed. The medium size SFR concepts are from the Advanced Burner Reactors (ABR). The large size SFR concepts are from the self-breeding reactors. The TRIPOLI-4 depletion calculations were made with MOX and metallic U-Pu-Zr fuels for the ABR cores and with MOX and Carbide (U,Pu)C fuels for the self-breeding cores. The whole core reactor physics parameters calculations were performed for the BOEC and EOEC (Beginning and End of Equilibrium Cycle) conditions. This paper summarizes the TRIPOLI-4 calculation results of Keff, βeff, sodium void worth, Doppler constant, control rod worth, and core power distributions for the BOEC and EOEC conditions. The one-cycle depletion calculation results of the core inventory of U and TRU (Pu, Am, Cm, and Np) are also analyzed, after 328.5 days depletion irradiation for the ABR cores, 410 days for the large MOX core, and 500 days for the large carbide core. (author)
Brogan, John
Understanding the dosimetry for high-energy, heavy ions (HZE), especially within living systems, is complex and requires the use of both experimental and computational methods. Tissue-equivalent proportional counters (TEPCs) have been used experimentally to measure energy deposition in volumes similar in dimension to a mammalian cell. As these experiments begin to include a wider range of ions and energies, considerations to cost, time, and radiation protection are necessary and may limit the extent of these studies. Multiple Monte Carlo computational codes have been created to remediate this problem and serve as a mode of verification for pervious experimental methods. One such code, Relativistic-Ion Tracks (RITRACKS), is currently being developed at the NASA Johnson Space center. RITRACKS was designed to describe patterns of ionizations responsible for DNA damage on the molecular scale (nanometers). This study extends RITRACKS version 3.07 into the microdosimetric scale (microns), and compares computational results to previous experimental TEPC data. Energy deposition measurements for 1000 MeV nucleon-1 Fe ions in a 1 micron spherical target were compared. Different settings within RITRACKS were tested to verify their effects on dose to a target and the resulting energy deposition frequency distribution. The results were then compared to the TEPC data.
Burnup calculation capability in the PSG2 / Serpent Monte Carlo reactor physics code
The PSG continuous-energy Monte Carlo reactor physics code has been developed at VTT Technical Research Centre of Finland since 2004. The code is mainly intended for group constant generation for coupled reactor simulator calculations and other tasks traditionally handled using deterministic lattices physics codes. The name was recently changed from acronym PSG to 'Serpent', and the capabilities have been extended by implementing built-in burnup calculation routines that enable the code to be used for fuel cycle studies and the modelling of irradiated fuels. This paper presents the methodology used for burnup calculation. Serpent has two fundamentally different options for solving the Bateman depletion equations: 1) the Transmutation Trajectory Analysis method (TTA), based on the analytical solution of linearized depletion chains and 2) the Chebyshev Rational Approximation Method (CRAM), an advanced matrix exponential solution developed at VTT. The first validation results are compared to deterministic CASMO-4E calculations. It is also shown that the overall running time in Monte Carlo burnup calculation can be significantly reduced using specialized calculation techniques, and that the continuous-energy Monte Carlo method is becoming a viable alternative to deterministic assembly burnup codes. (authors)
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...
Dose conversion coefficients for ICRP110 voxel phantom in the Geant4 Monte Carlo code
Martins, M. C.; Cordeiro, T. P. V.; Silva, A. X.; Souza-Santos, D.; Queiroz-Filho, P. P.; Hunt, J. G.
2014-02-01
The reference adult male voxel phantom recommended by International Commission on Radiological Protection no. 110 was implemented in the Geant4 Monte Carlo code. Geant4 was used to calculate Dose Conversion Coefficients (DCCs) expressed as dose deposited in organs per air kerma for photons, electrons and neutrons in the Annals of the ICRP. In this work the AP and PA irradiation geometries of the ICRP male phantom were simulated for the purpose of benchmarking the Geant4 code. Monoenergetic photons were simulated between 15 keV and 10 MeV and the results were compared with ICRP 110, the VMC Monte Carlo code and the literature data available, presenting a good agreement.
ERSN-OpenMC, a Java-based GUI for OpenMC Monte Carlo code
Jaafar EL Bakkali
2016-07-01
Full Text Available OpenMC is a new Monte Carlo transport particle simulation code focused on solving two types of neutronic problems mainly the k-eigenvalue criticality fission source problems and external fixed fission source problems. OpenMC does not have any Graphical User Interface and the creation of one is provided by our java-based application named ERSN-OpenMC. The main feature of this application is to provide to the users an easy-to-use and flexible graphical interface to build better and faster simulations, with less effort and great reliability. Additionally, this graphical tool was developed with several features, as the ability to automate the building process of OpenMC code and related libraries as well as the users are given the freedom to customize their installation of this Monte Carlo code. A full description of the ERSN-OpenMC application is presented in this paper.
A new Monte Carlo code for absorption simulation of laser-skin tissue interaction
Afshan Shirkavand; Saeed Sarkar; Marjaneh Hejazi; Leila Ataie-Fashtami; Mohammad Reza Alinaghizadeh
2007-01-01
In laser clinical applications, the process of photon absorption and thermal energy diffusion in the target tissue and its surrounding tissue during laser irradiation are crucial. Such information allows the selection of proper operating parameters such as laser power, and exposure time for optimal therapeutic. The Monte Carlo method is a useful tool for studying laser-tissue interaction and simulation of energy absorption in tissue during laser irradiation. We use the principles of this technique and write a new code with MATLAB 6.5, and then validate it against Monte Carlo multi layer (MCML) code. The new code is proved to be with good accuracy. It can be used to calculate the total power bsorbed in the region of interest. This can be combined for heat modelling with other computerized programs.
Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William
2004-06-01
ITS is a powerful and user-friendly software package permitting state of the art Monte Carlo solution of linear time-independent couple electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 5.0, the latest version of ITS, contains (1) improvements to the ITS 3.0 continuous-energy codes, (2)multigroup codes with adjoint transport capabilities, and (3) parallel implementations of all ITS codes. Moreover the general user friendliness of the software has been enhanced through increased internal error checking and improved code portability.
ITS version 5.0 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes
ITS is a powerful and user-friendly software package permitting state of the art Monte Carlo solution of linear time-independent couple electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 5.0, the latest version of ITS, contains (1) improvements to the ITS 3.0 continuous-energy codes, (2)multigroup codes with adjoint transport capabilities, and (3) parallel implementations of all ITS codes. Moreover the general user friendliness of the software has been enhanced through increased internal error checking and improved code portability
PyMercury: Interactive Python for the Mercury Monte Carlo Particle Transport Code
Iandola, F N; O' Brien, M J; Procassini, R J
2010-11-29
Monte Carlo particle transport applications are often written in low-level languages (C/C++) for optimal performance on clusters and supercomputers. However, this development approach often sacrifices straightforward usability and testing in the interest of fast application performance. To improve usability, some high-performance computing applications employ mixed-language programming with high-level and low-level languages. In this study, we consider the benefits of incorporating an interactive Python interface into a Monte Carlo application. With PyMercury, a new Python extension to the Mercury general-purpose Monte Carlo particle transport code, we improve application usability without diminishing performance. In two case studies, we illustrate how PyMercury improves usability and simplifies testing and validation in a Monte Carlo application. In short, PyMercury demonstrates the value of interactive Python for Monte Carlo particle transport applications. In the future, we expect interactive Python to play an increasingly significant role in Monte Carlo usage and testing.
PyMercury: Interactive Python for the Mercury Monte Carlo Particle Transport Code
Monte Carlo particle transport applications are often written in low-level languages (C/C++) for optimal performance on clusters and supercomputers. However, this development approach often sacrifices straightforward usability and testing in the interest of fast application performance. To improve usability, some high-performance computing applications employ mixed-language programming with high-level and low-level languages. In this study, we consider the benefits of incorporating an interactive Python interface into a Monte Carlo application. With PyMercury, a new Python extension to the Mercury general-purpose Monte Carlo particle transport code, we improve application usability without diminishing performance. In two case studies, we illustrate how PyMercury improves usability and simplifies testing and validation in a Monte Carlo application. In short, PyMercury demonstrates the value of interactive Python for Monte Carlo particle transport applications. In the future, we expect interactive Python to play an increasingly significant role in Monte Carlo usage and testing.
The use of an inbuilt importance generator for acceleration of the Monte Carlo code MCBEND
Monte Carlo is currently the most accurate method for the analysis of neutron and gamma-ray transport. However its application, especially to deep penetration studies, is costly in terms of the man-days to set up the calculation and in terms of computer usage. The MAGIC module, developed at the Winfrith Technology Centre, addresses both these problems. It employs an automated procedure based upon the established technique of splitting/roulette with an importance function derived from the solution of the adjoint diffusion equation. Examples are given of the application of the module with Monte Carlo code MCBEND
Estimation of skyshine dose from turbine building of BWR plant using Monte Carlo code
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)
A new assembly-level Monte Carlo neutron transport code for reactor physics calculations
This paper presents a new assembly-level Monte Carlo neutron transport code, specifically intended for diffusion code group-constant generation and other reactor physics calculations. The code is being developed at the Technical Research Centre of Finland (VTT), under the working title 'Probabilistic Scattering Game', or PSG. The PSG code uses a method known as Woodcock tracking to simulate neutron histories. The advantages of the method include fast simulation in complex geometries and relatively simple handling of complicated geometrical objects. The main drawback is the inability to calculate reaction rates in optically thin volumes. This narrows the field of application to calculations involving parameters integrated over large volumes. The main features of the PSG code and the Woodcock tracking method are introduced. The code is applied in three example cases, involving infinite lattices of two-dimensional LWR fuel assemblies. Comparison calculations are carried out using MCNP4C and CASMO-4E. The results reveal that the code performs quite well in the calculation cases of this study, especially when compared to MCNP. The PSG code is still under extensive development and there are both flaws in the simulation of the interaction physics and programming errors in the source code. The results presented here, however, seem very encouraging, especially considering the early development stage of the code. (author)
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...
Srna - Monte Carlo codes for proton transport simulation in combined and voxelized geometries
Ilić Radovan D.
2002-01-01
Full Text Available This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtained through the PETRA and GEANT programs. The simulation of the proton beam characterization by means of the Multi-Layer Faraday Cup and spatial distribution of positron emitters obtained by our program indicate the imminent application of Monte Carlo techniques in clinical practice.
Srna-Monte Carlo codes for proton transport simulation in combined and voxelized geometries
This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D) dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtained through the PETRA and GEANT programs. The simulation of the proton beam characterization by means of the Multi-Layer Faraday Cup and spatial distribution of positron emitters obtained by our program indicate the imminent application of Monte Carlo techniques in clinical practice. (author)
Calculation of effective delayed neutron fraction with modified library of Monte Carlo code
Highlights: ► We propose a new Monte Carlo method to calculate the effective delayed neutron fraction by changing the library. ► We study the stability of our method. When the particles and cycles are sufficiently great, the stability is very good. ► The final result is determined to make the deviation least. ► We verify our method on several benchmarks, and the results are very good. - Abstract: A new Monte Carlo method is proposed to calculate the effective delayed neutron fraction βeff. Based on perturbation theory, βeff is calculated with modified library of Monte Carlo code. To verify the proposed method, calculations are performed on several benchmarks. The error of the method is analyzed and the way to reduce error is proposed. The results are in good agreement with the reference data
Introduction to the simulation with MCNP Monte Carlo code and its applications in Medical Physics
The simulation by Monte Carlo is tool which Medical Physics counts with it for the development of its research, the interest by this tool is growing, as we may observe in the main scientific journals for the years 1995-1997 where more than 27 % of the papers treat over Monte Carlo and/or its applications in the radiation transport.In the Peruvian Institute of Nuclear Energy we are implementing and making use of the MCNP4 and EGS4 codes. In this work are presented the general features of the Monte Carlo method and its more useful applications in Medical Physics. Likewise, it is made a simulation of the calculation of isodose curves in an interstitial treatment with Ir-192 wires in a mammary gland carcinoma. (Author)
Recently several code systems have been developed and utilized at Japan Atomic Energy Research Institute. It is not easy to construct and maintain a code system, but the fact is not well known in the researchers. For this reason, it will be very useful to publish informations about design concepts, characteristics, necessary computer features and amounts of invested manpower for the developments of some recent JAERI code systems. In this report, a general view of required manpower on unification of nuclear codes is discussed and four code systems, i.e., SRAC for thermal reactor analysis, TRITON for Tokamak MHD analysis, SPEEDI for emergent environmental dose prediction and RADHEAT for radiation shielding analysis are presented. They are described from aspects of (1) purpose and schedule of development, (2) outline of system, (3) results of benchmark tests, (4) utilized computer features, (5) invested manpower, and (6) desirable computer features. Finally common aspects of four code systems from viewpoint of necessary computer hardwares and softwares are discussed for future development of code systems. (author)
DgSMC-B code: A robust and autonomous direct simulation Monte Carlo code for arbitrary geometries
Kargaran, H.; Minuchehr, A.; Zolfaghari, A.
2016-07-01
In this paper, we describe the structure of a new Direct Simulation Monte Carlo (DSMC) code that takes advantage of combinatorial geometry (CG) to simulate any rarefied gas flows Medias. The developed code, called DgSMC-B, has been written in FORTRAN90 language with capability of parallel processing using OpenMP framework. The DgSMC-B is capable of handling 3-dimensional (3D) geometries, which is created with first-and second-order surfaces. It performs independent particle tracking for the complex geometry without the intervention of mesh. In addition, it resolves the computational domain boundary and volume computing in border grids using hexahedral mesh. The developed code is robust and self-governing code, which does not use any separate code such as mesh generators. The results of six test cases have been presented to indicate its ability to deal with wide range of benchmark problems with sophisticated geometries such as airfoil NACA 0012. The DgSMC-B code demonstrates its performance and accuracy in a variety of problems. The results are found to be in good agreement with references and experimental data.
FAST CONVERGENT MONTE CARLO RECEIVER FOR OFDM SYSTEMS
Wu Lili; Liao Guisheng; Bao Zheng; Shang Yong
2005-01-01
The paper investigates the problem of the design of an optimal Orthogonal Frequency Division Multiplexing (OFDM) receiver against unknown frequency selective fading. A fast convergent Monte Carlo receiver is proposed. In the proposed method, the Markov Chain Monte Carlo (MCMC) methods are employed for the blind Bayesian detection without channel estimation. Meanwhile, with the exploitation of the characteristics of OFDM systems, two methods are employed to improve the convergence rate and enhance the efficiency of MCMC algorithms.One is the integration of the posterior distribution function with respect to the associated channel parameters, which is involved in the derivation of the objective distribution function; the other is the intra-symbol differential coding for the elimination of the bimodality problem resulting from the presence of unknown fading channels. Moreover, no matrix inversion is needed with the use of the orthogonality property of OFDM modulation and hence the computational load is significantly reduced. Computer simulation results show the effectiveness of the fast convergent Monte Carlo receiver.
On the use of SERPENT Monte Carlo code to generate few group diffusion constants
Piovezan, Pamela, E-mail: pamela.piovezan@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Carluccio, Thiago; Domingos, Douglas Borges; Rossi, Pedro Russo; Mura, Luiz Felipe, E-mail: fermium@cietec.org.b, E-mail: thiagoc@ipen.b [Fermium Tecnologia Nuclear, Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2011-07-01
The accuracy of diffusion reactor codes strongly depends on the quality of the groups constants processing. For many years, the generation of such constants was based on 1-D infinity cell transport calculations. Some developments using collision probability or the method of characteristics allow, nowadays, 2-D assembly group constants calculations. However, these 1-D and 2-D codes how some limitations as , for example, on complex geometries and in the neighborhood of heavy absorbers. On the other hand, since Monte Carlos (MC) codes provide accurate neutro flux distributions, the possibility of using these solutions to provide group constants to full-core reactor diffusion simulators has been recently investigated, especially for the cases in which the geometry and reactor types are beyond the capability of the conventional deterministic lattice codes. The two greatest difficulties on the use of MC codes to group constant generation are the computational costs and the methodological incompatibility between analog MC particle transport simulation and deterministic transport methods based in several approximations. The SERPENT code is a 3-D continuous energy MC transport code with built-in burnup capability that was specially optimized to generate these group constants. In this work, we present the preliminary results of using the SERPENT MC code to generate 3-D two-group diffusion constants for a PWR like assembly. These constants were used in the CITATION diffusion code to investigate the effects of the MC group constants determination on the neutron multiplication factor diffusion estimate. (author)
Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code
High-temperature gas-cooled reactor fuels are composed of thousands of microscopic fuel particles, randomly dispersed in a graphite matrix. The modelling of such geometry is complicated, especially using continuous-energy Monte Carlo codes, which are unable to apply any deterministic corrections in the calculation. This paper presents the geometry routine developed for modelling randomly dispersed particle fuels using the PSG Monte Carlo reactor physics code. The model is based on the delta-tracking method, and it takes into account the spatial self-shielding effects and the random dispersion of the fuel particles. The calculation routine is validated by comparing the results to reference MCNP4C calculations using uranium and plutonium based fuels. (authors)
Evaluation of CASMO-3 and HELIOS for Fuel Assembly Analysis from Monte Carlo Code
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.
TRIPOLI-4{sup ®} Monte Carlo code ITER A-lite neutronic model validation
Jaboulay, Jean-Charles, E-mail: jean-charles.jaboulay@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Cayla, Pierre-Yves; Fausser, Clement [MILLENNIUM, 16 Av du Québec Silic 628, F-91945 Villebon sur Yvette (France); Damian, Frederic; Lee, Yi-Kang; Puma, Antonella Li; Trama, Jean-Christophe [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France)
2014-10-15
3D Monte Carlo transport codes are extensively used in neutronic analysis, especially in radiation protection and shielding analyses for fission and fusion reactors. TRIPOLI-4{sup ®} is a Monte Carlo code developed by CEA. The aim of this paper is to show its capability to model a large-scale fusion reactor with complex neutron source and geometry. A benchmark between MCNP5 and TRIPOLI-4{sup ®}, on the ITER A-lite model was carried out; neutron flux, nuclear heating in the blankets and tritium production rate in the European TBMs were evaluated and compared. The methodology to build the TRIPOLI-4{sup ®} A-lite model is based on MCAM and the MCNP A-lite model. Simplified TBMs, from KIT, were integrated in the equatorial-port. A good agreement between MCNP and TRIPOLI-4{sup ®} is shown; discrepancies are mainly included in the statistical error.
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...
Shielding evaluation for e-Linac - Inter-comparison of Monte Carlo codes and analytical calculations
Estimation of optimum shielding thickness is an important aspect in radiation protection as well as in assessment of cost effectiveness of any upcoming accelerator facility. Analytical calculations for shielding estimates are fast and being frequently used even though they are very approximate. Estimates by Monte Carlo codes, on the other hand is accurate, provided used in a judicious manner, but they are very time consuming and require high end computational hardware. The purpose of this work is to compare the results from various available Monte Carlo codes, such as FLUKA and EGSmc. The estimated output was also compared with the analytical techniques. For the work, an e-Linac facility of 50 MeV electron beam was used and calculations were carried out with 1 mA beam current. (author)
COG is a major multiparticle simulation code in the LLNL Monte Carlo radiation transport toolkit. It was designed to solve deep-penetration radiation shielding problems in arbitrarily complex 3D geometries, involving coupled transport of photons, neutrons, and electrons. COG was written to provide as much accuracy as the underlying cross-sections will allow, and has a number of variance-reduction features to speed computations. Recently COG has been applied to the simulation of high- resolution radiographs of complex objects and the evaluation of contraband detection schemes. In this paper we will give a brief description of the capabilities of the COG transport code and show several examples of neutron and gamma-ray imaging simulations. Keywords: Monte Carlo, radiation transport, simulated radiography, nonintrusive inspection, neutron imaging
Vectorization and multitasking with a Monte-Carlo code for neutron transport problems
This paper summarizes two improvements of a Monte Carlo code by resorting to vectorization and multitasking techniques. After a short presentation of the physical problem to solve and a description of the main difficulties to produce an efficient coding, this paper introduces the vectorization principles employed and briefly describes how the vectorized algorithm works. Next, measured performances on CRAY 1S, CYBER 205 and CRAY X-MP are compared. The second part of this paper is devoted to multitasking technique. Starting from the standard multitasking tools available with FORTRAN on CRAY X-MP/4, a multitasked algorithm and its measured speed-ups are presented. In conclusion we prove that vector and parallel computers are a great opportunity for such Monte Carlo algorithms
PEREGRINE: An all-particle Monte Carlo code for radiation therapy
The goal of radiation therapy is to deliver a lethal dose to the tumor while minimizing the dose to normal tissues. To carry out this task, it is critical to calculate correctly the distribution of dose delivered. Monte Carlo transport methods have the potential to provide more accurate prediction of dose distributions than currently-used methods. PEREGRINE is a new Monte Carlo transport code developed at Lawrence Livermore National Laboratory for the specific purpose of modeling the effects of radiation therapy. PEREGRINE transports neutrons, photons, electrons, positrons, and heavy charged-particles, including protons, deuterons, tritons, helium-3, and alpha particles. This paper describes the PEREGRINE transport code and some preliminary results for clinically relevant materials and radiation sources
Development of Monte Carlo decay gamma-ray transport calculation system
In the DT fusion reactor, it is critical concern to evaluate the decay gamma-ray biological dose rates after the reactor shutdown exactly. In order to evaluate the decay gamma-ray biological dose rates exactly, three dimensional Monte Carlo decay gamma-ray transport calculation system have been developed by connecting the three dimensional Monte Carlo particle transport calculation code and the induced activity calculation code. The developed calculation system consists of the following four functions. (1) The operational neutron flux distribution is calculated by the three dimensional Monte Carlo particle transport calculation code. (2) The induced activities are calculated by the induced activity calculation code. (3) The decay gamma-ray source distribution is obtained from the induced activities. (4) The decay gamma-rays are generated by using the decay gamma-ray source distribution, and the decay gamma-ray transport calculation is conducted by the three dimensional Monte Carlo particle transport calculation code. In order to reduce the calculation time drastically, a biasing system for the decay gamma-ray source distribution has been developed, and the function is also included in the present system. In this paper, the outline and the detail of the system, and the execution example are reported. The evaluation for the effect of the biasing system is also reported. (author)
Diego Ferraro
2011-01-01
Full Text Available Monte Carlo neutron transport codes are usually used to perform criticality calculations and to solve shielding problems due to their capability to model complex systems without major approximations. However, these codes demand high computational resources. The improvement in computer capabilities leads to several new applications of Monte Carlo neutron transport codes. An interesting one is to use this method to perform cell-level fuel assembly calculations in order to obtain few group constants to be used on core calculations. In the present work the VTT recently developed Serpent v.1.1.7 cell-oriented neutronic calculation code is used to perform cell calculations of a theoretical BWR lattice benchmark with burnable poisons, and the main results are compared to reported ones and with calculations performed with Condor v.2.61, the INVAP's neutronic collision probability cell code.
The Serpent Monte Carlo Code: Status, Development and Applications in 2013
Leppänen, Jaakko; Pusa, Maria; Viitanen, Tuomas; Valtavirta, Ville; Kaltiaisenaho, Toni
2014-06-01
The Serpent Monte Carlo reactor physics burnup calculation code has been developed at VTT Technical Research Centre of Finland since 2004, and is currently used in 100 universities and research organizations around the world. This paper presents the brief history of the project, together with the currently available methods and capabilities and plans for future work. Typical user applications are introduced in the form of a summary review on Serpent-related publications over the past few years.
A model of a gamma sterilizer was built using the ITS/ACCEPT Monte Carlo code and verified through dosimetry. Individual dosimetry measurements in homogeneous material were pooled to represent larger bodies that could be simulated in a reasonable time. With the assumptions and simplifications described, dose predictions were within 2-5% of dosimetry. The model was used to simulate product movement through the sterilizer and to predict information useful for process optimization and facility design
Validation of GEANT4 Monte Carlo Simulation Code for 6 MV Varian Linac Photon Beam
The head of a clinical linear accelerator based on the manufacturer detailed information is simulated by using GEANT4. Percentage Depth Dose (PDD) and flatness symmetry (lateral dose profiles) in water phantom were evaluated. Comparisons between experimental and simulated data were carried out for two field sizes; 5 × 5, and 10 ×10 cm2. The obtained results indicated that GEANT4 code is a promising and validated Monte Carlo program for using in radiotherapy applications
TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticallity studies
TRIMARAN is developed for safety analysis of nuclar components containing fissionnable materials: shipping casks, storage and cooling pools, manufacture and reprocessing plants. It solves the transport equation by Monte Carlo method in general three dimensional geometry with multigroup P1 approximation. A special representation of cross sections and numbers has been developed in order to reduce considerably the computing cost and allow this three dimensional code to compete with standard numerical program used in parametric studies
Monte-Carlo Simulation for PDC-Based Optical CDMA System
FAHIM AZIZ UMRANI
2010-10-01
Full Text Available This paper presents the Monte-Carlo simulation of Optical CDMA (Code Division Multiple Access systems, and analyse its performance in terms of the BER (Bit Error Rate. The spreading sequence chosen for CDMA is Perfect Difference Codes. Furthermore, this paper derives the expressions of noise variances from first principles to calibrate the noise for both bipolar (electrical domain and unipolar (optical domain signalling required for Monte-Carlo simulation. The simulated results conform to the theory and show that the receiver gain mismatch and splitter loss at the transceiver degrades the system performance.
Perfetti, C.; Martin, W. [Univ. of Michigan, Dept. of Nuclear Engineering and Radiological Sciences, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109-2104 (United States); Rearden, B.; Williams, M. [Oak Ridge National Laboratory, Reactor and Nuclear Systems Div., Bldg. 5700, P.O. Box 2008, Oak Ridge, TN 37831-6170 (United States)
2012-07-01
Three methods for calculating continuous-energy eigenvalue sensitivity coefficients were developed and implemented into the Shift Monte Carlo code within the SCALE code package. The methods were used for two small-scale test problems and were evaluated in terms of speed, accuracy, efficiency, and memory requirements. A promising new method for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was developed and produced accurate sensitivity coefficients with figures of merit that were several orders of magnitude larger than those from existing methods. (authors)
Exact modeling of the torus geometry with Monte Carlo transport code
It is valuable to model torus geometry exactry for the neutronics design of fusion reactor in order to assess neutronics characteristics such as tritium breeding ratio, heat generation rate, etc, near the plasma. Monte Carlo code MORSE-GG which plays important role in the radiation streaming calculation of fusion reactors had been able to deal with the geometry composed of second order surfaces. The MORSE-GG program is modified to be able to deal with torus geometry which has fourth order surface by solving biquadratic equations, hoping that MORSE-GG code becomes more effective for the neutronics calculation of the Tokamak fusion reactor. (author)
Efficient data management techniques implemented in the Karlsruhe Monte Carlo code KAMCCO
The Karlsruhe Monte Carlo Code KAMCCO is a forward neutron transport code with an eigenfunction and a fixed source option, including time-dependence. A continuous energy model is combined with a detailed representation of neutron cross sections, based on linear interpolation, Breit-Wigner resonances and probability tables. All input is processed into densely packed, dynamically addressed parameter fields and networks of pointers (addresses). Estimation routines are decoupled from random walk and analyze a storage region with sample records. This technique leads to fast execution with moderate storage requirements and without any I/O-operations except in the input and output stages. 7 references. (U.S.)
Procassini, R.J. [Lawrence Livermore National lab., CA (United States)
1997-12-31
The fine-scale, multi-space resolution that is envisioned for accurate simulations of complex weapons systems in three spatial dimensions implies flop-rate and memory-storage requirements that will only be obtained in the near future through the use of parallel computational techniques. Since the Monte Carlo transport models in these simulations usually stress both of these computational resources, they are prime candidates for parallelization. The MONACO Monte Carlo transport package, which is currently under development at LLNL, will utilize two types of parallelism within the context of a multi-physics design code: decomposition of the spatial domain across processors (spatial parallelism) and distribution of particles in a given spatial subdomain across additional processors (particle parallelism). This implementation of the package will utilize explicit data communication between domains (message passing). Such a parallel implementation of a Monte Carlo transport model will result in non-deterministic communication patterns. The communication of particles between subdomains during a Monte Carlo time step may require a significant level of effort to achieve a high parallel efficiency.
ASCOT: redesigned Monte Carlo code for simulations of minority species in tokamak plasmas
Hirvijoki, Eero; Koskela, Tuomas; Kurki-Suonio, Taina; Miettunen, Juho; Sipilä, Seppo; Snicker, Antti; Äkäslompolo, Simppa
2013-01-01
A comprehensive description of methods for Monte Carlo studies of fast ions and impurity species in tokamak plasmas is presented. The described methods include Hamiltonian orbit-following in particle and guiding center phase space, test particle or guiding center solution of the kinetic equation applying stochastic differential equations in the presence of Coulomb collisions, Neoclassical tearing modes and Alfv\\'en eigenmodes as electromagnetic perturbations relevant for fast ions, together with plasma flow and atomic reactions relevant for impurity studies. Applying the methods, a complete reimplementation of a well-established minority species code is carried out as a response both to the increase in computing power during the last twenty years and to the weakly structured growth of the previous code which has made implementation of additional models impractical. Also, a thorough benchmark between the previous code and the reimplementation is accomplished, showing good agreement between the codes.
Uncertainties associated with the use of the KENO Monte Carlo criticality codes
The KENO multi-group Monte Carlo criticality codes have earned the reputation of being efficient, user friendly tools especially suited for the analysis of situations commonly encountered in the storage and transportation of fissile materials. Throughout their twenty years of service, a continuing effort has been made to maintain and improve these codes to meet the needs of the nuclear criticality safety community. Foremost among these needs is the knowledge of how to utilize the results safely and effectively. Therefore it is important that code users be aware of uncertainties that may affect their results. These uncertainties originate from approximations in the problem data, methods used to process cross sections, and assumptions, limitations and approximations within the criticality computer code itself. 6 refs., 8 figs., 1 tab
Specific Monte Carlo code development for nuclear well-logging tool responses
McPNL is a specific Monte Carlo computer code that has been developed to simulate a pulsed neutron oil well logging tool and uses implicit capture, Russian roulette and statistical estimation techniques as primary variance reduction methods. The code has been validated by benchmarking against six sets of laboratory test pit data on water, limestone and quartz formations with widely varying sets of borehole and formation conditions. McDNL is a specific Monte Carlo computer code that has been developed to simulate a dual-spaced neutron porosity tool. The low counting yield in the far detector of the tool requires the use of biasing schemes to obtain adequate efficiency. Exponential transform and directional biasing techniques have been applied with remarkable success for this problem, along with source biasing, implicit capture, Russian roulette and statistical estimation techniques. The code has been benchmarked against five sets of laboratory test pit data and found to be valid. Correlated sampling can be optionally used in the code to accurately predict the relative change in the detector response due to small perturbations in the formation porosity. (author)
Assessment of MIRD data for internal dosimetry using the GATE Monte Carlo code.
Parach, Ali Asghar; Rajabi, Hossein; Askari, Mohammad Ali
2011-08-01
GATE/GEANT is a Monte Carlo code dedicated to nuclear medicine that allows calculation of the dose to organs of voxel phantoms. On the other hand, MIRD is a well-developed system for estimation of the dose to human organs. In this study, results obtained from GATE/GEANT using Snyder phantom are compared to published MIRD data. For this, the mathematical Snyder phantom was discretized and converted to a digital phantom of 100 × 200 × 360 voxels. The activity was considered uniformly distributed within kidneys, liver, lungs, pancreas, spleen, and adrenals. The GATE/GEANT Monte Carlo code was used to calculate the dose to the organs of the phantom from mono-energetic photons of 10, 15, 20, 30, 50, 100, 200, 500, and 1000 keV. The dose was converted into specific absorbed fraction (SAF) and the results were compared to the corresponding published MIRD data. On average, there was a good correlation (r (2)>0.99) between the two series of data. However, the GATE/GEANT data were on average -0.16 ± 6.22% lower than the corresponding MIRD data for self-absorption. Self-absorption in the lungs was considerably higher in the MIRD compared to the GATE/GEANT data, for photon energies of 10-20 keV. As for cross-irradiation to other organs, the GATE/GEANT data were on average +1.5 ± 8.1% higher than the MIRD data, for photon energies of 50-1000 keV. For photon energies of 10-30 keV, the relative difference was +7.5 ± 67%. It turned out that the agreement between the GATE/GEANT and the MIRD data depended upon absolute SAF values and photon energy. For 10-30 keV photons, where the absolute SAF values were small, the uncertainty was high and the effect of cross-section prominent, and there was no agreement between the GATE/GEANT results and the MIRD data. However, for photons of 50-1,000 keV, the bias was negligible and the agreement was acceptable. PMID:21573984
An analytical solution to a simplified EDXRF model for Monte Carlo code verification
The objective of this study is to obtain an analytical solution to the scalar photon transport equation that can be used to obtain benchmark results for the verification of energy dispersive X-Ray fluorescence (EDXRF) Monte Carlo simulation codes. The multi-collided flux method (multiple scattering method) is implemented to obtain analytical expressions for the space-, energy-, and angle-dependent scalar photon flux for a one dimensional EDXRF model problem. In order to obtain benchmark results, higher-order multiple scattering terms are included in the multi-collided flux method. The details of the analytical solution and of the proposed EDXRF model problem are presented. Analytical expressions obtained are then used to calculate the energy-dependent current. The analytically-calculated energy-dependent current is compared with Monte Carlo code results. The findings of this study show that analytical solutions to the scalar photon transport equation with the proposed model problem can be used as a verification tool in EDXRF Monte Carlo code development.
This paper presents an unstructured mesh based multi-physics interface implemented in the Serpent 2 Monte Carlo code, for the purpose of coupling the neutronics solution to component-scale thermal hydraulics calculations, such as computational fluid dynamics (CFD). The work continues the development of a multi-physics coupling scheme, which relies on the separation of state-point information from the geometry input, and the capability to handle temperature and density distributions by a rejection sampling algorithm. The new interface type is demonstrated by a simplified molten-salt reactor test case, using a thermal hydraulics solution provided by the CFD solver in OpenFOAM. (author)
High energy photon (over 10 MeV) and neutron beams adopted in radiobiology and radiotherapy always produce mixed neutron/gamma-ray fields. The Mg(Ar) ionization chambers are commonly applied to determine the gamma-ray dose because of its neutron insensitive characteristic. Nowadays, many perturbation corrections for accurate dose estimation and lots of treatment planning systems are based on Monte Carlo technique. The Monte Carlo codes EGSnrc, FLUKA, GEANT4, MCNP5, and MCNPX were used to evaluate energy dependent response functions of the Exradin M2 Mg(Ar) ionization chamber to a parallel photon beam with mono-energies from 20 keV to 20 MeV. For the sake of validation, measurements were carefully performed in well-defined (a) primary M-100 X-ray calibration field, (b) primary 60Co calibration beam, (c) 6-MV, and (d) 10-MV therapeutic beams in hospital. At energy region below 100 keV, MCNP5 and MCNPX both had lower responses than other codes. For energies above 1 MeV, the MCNP ITS-mode greatly resembled other three codes and the differences were within 5%. Comparing to the measured currents, MCNP5 and MCNPX using ITS-mode had perfect agreement with the 60Co, and 10-MV beams. But at X-ray energy region, the derivations reached 17%. This work shows us a better insight into the performance of different Monte Carlo codes in photon-electron transport calculation. Regarding the application of the mixed field dosimetry like BNCT, MCNP with ITS-mode is recognized as the most suitable tool by this work.
Evaluation of Monte Carlo Codes Regarding the Calculated Detector Response Function in NDP Method
Tuan, Hoang Sy Minh; Sun, Gwang Min; Park, Byung Gun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2014-10-15
The basis of the NDP is the irradiation of a sample with a thermal or cold neutron beam and the subsequent release of charged particles due to neutron-induced exoergic charged particle reactions. Neutrons interact with the nuclei of elements and release mono-energetic charged particles, e.g. alpha particles or protons, and recoil atoms. Depth profile of the analyzed element can be obtained by making a linear transformation of the measured energy spectrum by using the stopping power of the sample material. A few micrometer of the material can be analyzed nondestructively, and on the order of 10nm depth resolution can be obtained depending on the material type with NDP method. In the NDP method, the one first steps of the analytical process is a channel-energy calibration. This calibration is normally made with the experimental measurement of NIST Standard Reference Material sample (SRM-93a). In this study, some Monte Carlo (MC) codes were tried to calculate the Si detector response function when this detector accounted the energy charges particles emitting from an analytical sample. In addition, these MC codes were also tried to calculate the depth distributions of some light elements ({sup 10}B, {sup 3}He, {sup 6}Li, etc.) in SRM-93a and SRM-2137 samples. These calculated profiles were compared with the experimental profiles and SIMS profiles. In this study, some popular MC neutron transport codes are tried and tested to calculate the detector response function in the NDP method. The simulations were modeled based on the real CN-NDP system which is a part of Cold Neutron Activation Station (CONAS) at HANARO (KAERI). The MC simulations are very successful at predicting the alpha peaks in the measured energy spectrum. The net area difference between the measured and predicted alpha peaks are less than 1%. A possible explanation might be bad cross section data set usage in the MC codes for the transport of low energetic lithium atoms inside the silicon substrate.