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)
Sanchez, R.A.; Fernandez V, J.M.; Salvat, F. [Servicio de Oncologia Radioterapica. Hospital Clinico de Barcelona. Villarroel 170 08036 Barcelona (Spain)
1998-12-31
In the present communication it is presented the results of the simulation utilizing the Penelope code (Penetration and Energy loss of Positrons and Electrons) in several applications of radiotherapy which can be the radioactive sources simulation: {sup 192} Ir, {sup 125} I, {sup 106} Ru or the electron beams simulation of a linear accelerator Siemens KDS. The simulations presented in this communication have been on computers of type Pentium PC of 100 throughout 300 MHz, and the times of execution were from some hours until several days depending of the complexity of the problem. It is concluded that Penelope is a very useful tool for the Monte Carlo calculations due to its great ability and its relative handling facilities. (Author)
RodrIguez, M L [Centro Medico Paitilla. Calle 53 y ave Balboa, Paitilla (Panama)], E-mail: milrocas@gmail.com
2008-09-07
In this work we present PENLINAC, a code package developed to facilitate the use of the Monte Carlo code PENELOPE for the simulation of therapeutic beams, including high-energy electrons, photons and {sup 60}Co beams. The code simplifies the creation of the treatment machine geometry, allowing the modeling of their components from elementary geometric bodies and their further conversion to the quadric functions-based structure handled by PENELOPE. The code is implemented in various subroutines that allow the user to handle several models of radiation sources and phase spaces. The phase spaces are not part of the geometry and can store many variables of the particle in a relatively small data space. The set of subroutines does not alter the PENELOPE algorithms; thus, the main program implemented by the user can maintain its kind-of-particle-independent structure. A support program can handle and analyze the phase spaces to generate, among others, last interaction maps and probability distributions that can be used as sources in simulation. Results from simulations of a Clinac linear accelerator head are presented in order to demonstrate the package capabilities. Dose distributions calculated in a water phantom for a variety of beams of this accelerator showed good agreement with measurements.
Application of a Monte Carlo Penelope code at diverse dosimetric problems in radiotherapy
In the present communication it is presented the results of the simulation utilizing the Penelope code (Penetration and Energy loss of Positrons and Electrons) in several applications of radiotherapy which can be the radioactive sources simulation: 192 Ir, 125 I, 106 Ru or the electron beams simulation of a linear accelerator Siemens KDS. The simulations presented in this communication have been on computers of type Pentium PC of 100 throughout 300 MHz, and the times of execution were from some hours until several days depending of the complexity of the problem. It is concluded that Penelope is a very useful tool for the Monte Carlo calculations due to its great ability and its relative handling facilities. (Author)
Simulation of clinical X-ray tube using the Monte Carlo Method - PENELOPE code
Breast cancer is the most common type of cancer among women. The main strategy to increase the long-term survival of patients with this disease is the early detection of the tumor, and mammography is the most appropriate method for this purpose. Despite the reduction of cancer deaths, there is a big concern about the damage caused by the ionizing radiation to the breast tissue. To evaluate these measures it was modeled a mammography equipment, and obtained the depth spectra using the Monte Carlo method - PENELOPE code. The average energies of the spectra in depth and the half value layer of the mammography output spectrum. (author)
Accurate simulation of ionization chamber response with the Monte Carlo code PENELOPE
Sempau, Josep [Technical University of Catalonia (Spain)
2010-07-01
Full text. Ionization chambers (IC) are routinely used in hospitals for the dosimetry of the photon and electron beams used for radiotherapy treatments. The determination of absorbed dose to water from the absorbed dose to the air filling the cavity requires the introduction of stopping power ratios and perturbation factors, which account for the disturbance caused by the presence of the chamber. Although this may seem a problem readily amenable to Monte Carlo simulation, the fact is that the accurate determination of IC response has been, during the last 20 years, one of the most important challenges of the simulation of electromagnetic showers. The main difficulty stems from the use of condensed history techniques for electron and positron transport. This approach, which involves grouping a large number of interactions into a single artificial event, is known to produce the so-called interface effects when particles travel across surfaces separating different media. These effects are extremely important when the electron step length is not negligible compared to the size of the region being crossed, as it is the case with the cavity of an IC. The artifact, which becomes apparent when the chamber response shows a marked dependence on the adopted step size, can be palliated with the use of sophisticated electron transport algorithms. These topics will be discussed in the context of the transport model implemented in the Penelope code. The degree of violation of the Fano theorem for a simple, planar geometry, will be used as a measure of the stability of the algorithm with respect to variations of the electron step length, thus assessing the 'quality' of its condensed history scheme. It will be shown that, with a suitable choice of transport parameters, Penelope can simulate IC response with an accuracy of the order of 0.1%. (author)
Accurate simulation of ionisation chamber response with the Monte Carlo code PENELOPE
Ionisation chambers (IC) are routinely used in hospitals for the dosimetry of the photon and electron beams used for radiotherapy treatments. The determination of absorbed dose to water from the absorbed dose to the air filling the cavity requires the introduction of stopping power ratios and perturbation factors, which account for the disturbance caused by the presence of the chamber. Although this may seem a problem readily amenable to Monte Carlo simulation, the fact is that the accurate determination of IC response has been, for various decades, one of the most important challenges of the simulation of electromagnetic showers. The main difficulty stems from the use of condensed history techniques for electron and positron transport. This approach, which involves grouping a large number of interactions into a single artificial event, is known to produce the so-called interface effects when particles travel across surfaces separating different media. These effects can be sizeable when the electron step length is not negligible compared to the size of the region being crossed, as it is the case with the cavity of an IC. The artefact, which becomes apparent when the chamber response shows a marked dependence on the adopted step size, can be palliated with the use of sophisticated electron transport algorithms. These topics are discussed in the context of the transport model implemented in the PENELOPE code. The degree of violation of the Fano theorem for a simple, planar geometry, is used as a measure of the stability of the algorithm with respect to variations of the electron step length, thus assessing the 'quality' of its condensed history scheme. It is shown that, with a suitable choice of transport parameters, PENELOPE simulates IC response with an accuracy of the order of 0.1%.
Accurate simulation of ionization chamber response with the Monte Carlo code PENELOPE
Full text. Ionization chambers (IC) are routinely used in hospitals for the dosimetry of the photon and electron beams used for radiotherapy treatments. The determination of absorbed dose to water from the absorbed dose to the air filling the cavity requires the introduction of stopping power ratios and perturbation factors, which account for the disturbance caused by the presence of the chamber. Although this may seem a problem readily amenable to Monte Carlo simulation, the fact is that the accurate determination of IC response has been, during the last 20 years, one of the most important challenges of the simulation of electromagnetic showers. The main difficulty stems from the use of condensed history techniques for electron and positron transport. This approach, which involves grouping a large number of interactions into a single artificial event, is known to produce the so-called interface effects when particles travel across surfaces separating different media. These effects are extremely important when the electron step length is not negligible compared to the size of the region being crossed, as it is the case with the cavity of an IC. The artifact, which becomes apparent when the chamber response shows a marked dependence on the adopted step size, can be palliated with the use of sophisticated electron transport algorithms. These topics will be discussed in the context of the transport model implemented in the Penelope code. The degree of violation of the Fano theorem for a simple, planar geometry, will be used as a measure of the stability of the algorithm with respect to variations of the electron step length, thus assessing the 'quality' of its condensed history scheme. It will be shown that, with a suitable choice of transport parameters, Penelope can simulate IC response with an accuracy of the order of 0.1%. (author)
Electron absorbed dose comparison between MCNP5 and Penelope Monte Carlo code for microdosimetry
The objective of the present work was to compare electron absorbed dose results between two widespread used codes in international scientific community: MCNP5 and Penelope-2003. Individual water spheres with masses between 10-9 g up to 10-3 g immersed in an infinite water medium (density of 1g/cm3) and monoenergetic electron sources with energy from 0.002 MeV to 0.1 MeV have been considered. The absorbed dose in the spheres was evaluated by both codes and the relative differences have been quantified. The results shown that Penelope gives, in general, higher results that, in some cases saturate or reach a maximum point and then rapidly drops. Particularly, for the 40 keV electron source we have done additional tests in three different scenarios: more points in the region of lower masses to a better definition of the curve behavior; MCNP used 200 substeps and Penelope was set to a full detail history methodology, and almost same parameters of case B but with the density of exterior medium increased to 10 g/cm3. The three cases show the influence of the backscattering that contribute with an important fraction of absorbed dose, finally we can infer a range of reliability to use the codes in this kind of simulations: both codes can calculate close results for up to 10-4 g.Even though MCNP5 uses the condensed history method, if simulation parameters are chosen carefully it can reproduce results very close to those obtained using detailed history mode. In some cases, the use of higher number of electron substeps causes significant differences in the result. (author)
Torres, J; Almansa, J F; Guerrero, R; Lallena, A M; Torres, Javier; Buades, Manuel J.; Almansa, Julio F.; Guerrero, Rafael; Lallena, Antonio M.
2003-01-01
Monte Carlo calculations using the codes PENELOPE and GEANT4 have been performed to characterize the dosimetric parameters of the new 20 mm long catheter based $^{32}$P beta source manufactured by Guidant Corporation. The dose distribution along the transverse axis and the two dimensional dose rate table have been calculated. Also, the dose rate at the reference point, the radial dose function and the anisotropy function were evaluated according to the adapted TG-60 formalism for cylindrical sources. PENELOPE and GEANT4 codes were first verified against previous results corresponding to the old 27 mm Guidant $^{32}$P beta source. The dose rate at the reference point for the unsheathed 27 mm source in water was calculated to be $0.215 \\pm 0.001$ cGy s$^{-1}$ mCi$^{-1}$, for PENELOPE, and $0.2312 \\pm 0.0008$ cGy s$^{-1}$ mCi$^{-1}$, for GEANT4. For the unsheathed 20 mm source these values were $0.2908 \\pm 0.0009$ cGy s$^{-1}$ mCi$^{-1}$ and $0.311 \\pm 0.001$ cGy s$^{-1}$ mCi$^{-1}$, respectively. Also, a compar...
PENELOPE, an algorithm and computer code for Monte Carlo simulation of electron-photon showers
Salvat, F.; Fernandez-Varea, J.M.; Baro, J.; Sempau, J.
1996-07-01
The FORTRAN 77 subroutine package PENELOPE performs Monte Carlo simulation of electron-photon showers in arbitrary for a wide energy range, from 1 keV to several hundred MeV. Photon transport is simulated by means of the standard, detailed simulation scheme. Electron and positron histories are generated on the basis of a mixed procedure, which combines detailed simulation of hard events with condensed simulation of soft interactions. A simple geometry package permits the generation of random electron-photon showers in material systems consisting of homogeneous bodies limited by quadric surfaces, i.e. planes, spheres, cylinders, etc. This report is intended not only to serve as a manual of the simulation package, but also to provide the user with the necessary information to understand the details of the Monte Carlo algorithm. (Author) 108 refs.
PENELOPE, an algorithm and computer code for Monte Carlo simulation of electron-photon showers
The FORTRAN 77 subroutine package PENELOPE performs Monte Carlo simulation of electron-photon showers in arbitrary for a wide energy range, from 1 keV to several hundred MeV. Photon transport is simulated by means of the standard, detailed simulation scheme. Electron and positron histories are generated on the basis of a mixed procedure, which combines detailed simulation of hard events with condensed simulation of soft interactions. A simple geometry package permits the generation of random electron-photon showers in material systems consisting of homogeneous bodies limited by quadric surfaces, i.e. planes, spheres, cylinders, etc. This report is intended not only to serve as a manual of the simulation package, but also to provide the user with the necessary information to understand the details of the Monte Carlo algorithm. (Author) 108 refs
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.)
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.)
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 work we have developed a simulation tool, based on the PENELOPE code, to study the response of MOSFET devices to irradiation with high-energy photons. The energy deposited in the extremely thin silicon dioxide layer has been calculated. To reduce the statistical uncertainties, an ant colony algorithm has been implemented to drive the application of splitting and Russian roulette as variance reduction techniques. In this way, the uncertainty has been reduced by a factor of ∼5, while the efficiency is increased by a factor of above 20. As an application, we have studied the dependence of the response of the pMOS transistor 3N163, used as a dosimeter, with the incidence angle of the radiation for three common photons sources used in radiotherapy: a 60Co Theratron-780 and the 6 and 18 MV beams produced by a Mevatron KDS LINAC. Experimental and simulated results have been obtained for gantry angles of 0 deg., 15 deg., 30 deg., 45 deg., 60 deg. and 75 deg. The agreement obtained has permitted validation of the simulation tool. We have studied how to reduce the angular dependence of the MOSFET response by using an additional encapsulation made of brass in the case of the two LINAC qualities considered.
Carvajal, M A; Palma, A J [Departamento de Electronica y Tecnologia de Computadores, Universidad de Granada, E-18071 Granada (Spain); Garcia-Pareja, S [Servicio de Radiofisica Hospitalaria, Hospital Regional Universitario ' Carlos Haya' , Avda Carlos Haya, s/n, E-29010 Malaga (Spain); Guirado, D [Servicio de RadiofIsica, Hospital Universitario ' San Cecilio' , Avda Dr Oloriz, 16, E-18012 Granada (Spain); Vilches, M [Servicio de Fisica y Proteccion Radiologica, Hospital Regional Universitario ' Virgen de las Nieves' , Avda Fuerzas Armadas, 2, E-18014 Granada (Spain); Anguiano, M; Lallena, A M [Departamento de Fisica Atomica, Molecular y Nuclear, Universidad de Granada, E-18071 Granada (Spain)], E-mail: carvajal@ugr.es, E-mail: garciapareja@gmail.com, E-mail: dguirado@ugr.es, E-mail: mvilches@ugr.es, E-mail: mangui@ugr.es, E-mail: ajpalma@ugr.es, E-mail: lallena@ugr.es
2009-10-21
In this work we have developed a simulation tool, based on the PENELOPE code, to study the response of MOSFET devices to irradiation with high-energy photons. The energy deposited in the extremely thin silicon dioxide layer has been calculated. To reduce the statistical uncertainties, an ant colony algorithm has been implemented to drive the application of splitting and Russian roulette as variance reduction techniques. In this way, the uncertainty has been reduced by a factor of {approx}5, while the efficiency is increased by a factor of above 20. As an application, we have studied the dependence of the response of the pMOS transistor 3N163, used as a dosimeter, with the incidence angle of the radiation for three common photons sources used in radiotherapy: a {sup 60}Co Theratron-780 and the 6 and 18 MV beams produced by a Mevatron KDS LINAC. Experimental and simulated results have been obtained for gantry angles of 0 deg., 15 deg., 30 deg., 45 deg., 60 deg. and 75 deg. The agreement obtained has permitted validation of the simulation tool. We have studied how to reduce the angular dependence of the MOSFET response by using an additional encapsulation made of brass in the case of the two LINAC qualities considered.
Carvajal, M A; García-Pareja, S; Guirado, D; Vilches, M; Anguiano, M; Palma, A J; Lallena, A M
2009-10-21
In this work we have developed a simulation tool, based on the PENELOPE code, to study the response of MOSFET devices to irradiation with high-energy photons. The energy deposited in the extremely thin silicon dioxide layer has been calculated. To reduce the statistical uncertainties, an ant colony algorithm has been implemented to drive the application of splitting and Russian roulette as variance reduction techniques. In this way, the uncertainty has been reduced by a factor of approximately 5, while the efficiency is increased by a factor of above 20. As an application, we have studied the dependence of the response of the pMOS transistor 3N163, used as a dosimeter, with the incidence angle of the radiation for three common photons sources used in radiotherapy: a (60)Co Theratron-780 and the 6 and 18 MV beams produced by a Mevatron KDS LINAC. Experimental and simulated results have been obtained for gantry angles of 0 degrees, 15 degrees, 30 degrees, 45 degrees, 60 degrees and 75 degrees. The agreement obtained has permitted validation of the simulation tool. We have studied how to reduce the angular dependence of the MOSFET response by using an additional encapsulation made of brass in the case of the two LINAC qualities considered. PMID:19794247
Were directly determined correction factors depending on the type camera beam quality, k, Q, and kQ, Qo, instead of the product (w, air p) Q, for three type cylindrical ionization chambers Pinpoint and divergent monoenergetic beams of photons in a wide range of energies (4-20 MV). The method of calculation used dispenses with the approaches taken in the classic procedure considered independent of braking power ratios and the factors disturbance of the camera. A detailed description of the geometry and materials chambers were supplied by the manufacturer and used as data input for the system 2006 of PENELOPE Monte Carlo calculation using a User code that includes correlated sampling, and forced interactions division of particles. We used a photon beam Co-60 as beam reference for calculating the correction factors for beam quality. No data exist for the cameras PTW 31014, 31015 and 31016 in the TRS-398 at they do not compare the results with data calculated or determined experimentally by other authors. (author)
Kahraman, A.; Kaya, S.; Jaksic, A.; Yilmaz, E.
2015-05-01
Radiation-sensing Field Effect Transistors (RadFETs or MOSFET dosimeters) with SiO2 gate dielectric have found applications in space, radiotherapy clinics, and high-energy physics laboratories. More sensitive RadFETs, which require modifications in device design, including gate dielectric, are being considered for personal dosimetry applications. This paper presents results of a detailed study of the RadFET energy response simulated with PENELOPE Monte Carlo code. Alternative materials to SiO2 were investigated to develop high-efficiency new radiation sensors. Namely, in addition to SiO2, Al2O3 and HfO2 were simulated as gate material and deposited energy amounts in these layers were determined for photon irradiation with energies between 20 keV and 5 MeV. The simulations were performed for capped and uncapped configurations of devices irradiated by point and extended sources, the surface area of which is the same with that of the RadFETs. Energy distributions of transmitted and backscattered photons were estimated using impact detectors to provide information about particle fluxes within the geometrical structures. The absorbed energy values in the RadFETs material zones were recorded. For photons with low and medium energies, the physical processes that affect the absorbed energy values in different gate materials are discussed on the basis of modelling results. The results show that HfO2 is the most promising of the simulated gate materials.
The authors report calculations performed using the MNCP and PENELOPE codes to determine the Hp(3)/K air conversion coefficient which allows the Hp(3) dose equivalent to be determined from the measured value of the kerma in the air. They report the definition of the phantom, a 20 cm diameter and 20 cm high cylinder which is considered as representative of a head. Calculations are performed for an energy range corresponding to interventional radiology or cardiology (20 keV-110 keV). Results obtained with both codes are compared
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)
This work has been performed within the frame of the European Union ORAMED project (Optimisation of Radiation protection for Medical staff). The main goal of the project is to improve standards of protection for medical staff for procedures resulting in potentially high exposures and to develop methodologies for better assessing and for reducing, exposures to medical staff. The Work Package WP2 is involved in the development of practical eye-lens dosimetry in interventional radiology. This study is complementary of the part of the ENEA report concerning the calculations with the MCNP-4C code of the conversion factors related to the operational quantity Hp(3). In this study, a set of energy- and angular-dependent conversion coefficients (Hp(3)/Ka), in the newly proposed square cylindrical phantom made of ICRU tissue, have been calculated with the Monte-Carlo code PENELOPE and MCNP5. The Hp(3) values have been determined in terms of absorbed dose, according to the definition of this quantity, and also with the kerma approximation as formerly reported in ICRU reports. At a low-photon energy (up to 1 MeV), the two results obtained with the two methods are consistent. Nevertheless, large differences are showed at a higher energy. This is mainly due to the lack of electronic equilibrium, especially for small angle incidences. The values of the conversion coefficients obtained with the MCNP-4C code published by ENEA quite agree with the kerma approximation calculations obtained with PENELOPE. We also performed the same calculations with the code MCNP5 with two types of tallies: F6 for kerma approximation and *F8 for estimating the absorbed dose that is, as known, due to secondary electrons. PENELOPE and MCNP5 results agree for the kerma approximation and for the absorbed dose calculation of Hp(3) and prove that, for photon energies larger than 1 MeV, the transport of the secondary electrons has to be taken into account. (authors)
The IEC 61267 (2005) specifies the filters used for the standard radiation conditions RQR-M, RQA-M, RQN-M and RQB-M. The filter has a thickness of (0.032 ± 0.002) mm, placed following the X-ray tube window. The photons spectra are generated by MC simulation for the Panallitical model PW-2185/00 X-ray tube between the voltages: 25 kV to 35 kV. Those spectra were compared with others references like the Catalogue of Diagnostic X-ray Spectra and Other Data (1997). The MC PENELOPE code was used to generate the photon spectra and the energy deposition calculations. (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)
The PENELOPE code system. Specific features and recent improvements
Since its first release, back in 1996, the Monte Carlo code system PENELOPE has evolved into a flexible and reliable tool for describing coupled electron-photon transport in complex material structures. The present article contains an overview of the physical interaction models, particle tracking methods, geometry tools, and variance-reduction techniques implemented in PENELOPE. Recent refinements aimed at improving the accuracy of the code, and its stability under variations of user-defined simulation parameters, are also described. These include the use of reliable cross sections for the ionization of inner atomic electron shells by electron/positron impact, a reformulation of the random-hinge method, and the use of fuzzy quadric surfaces in the description of the geometry. (author)
Chabert, I.; Barat, E.; Dautremer, T.; Montagu, T.; Agelou, M.; Croc de Suray, A.; Garcia-Hernandez, J. C.; Gempp, S.; Benkreira, M.; de Carlan, L.; Lazaro, D.
2016-07-01
This work aims at developing a generic virtual source model (VSM) preserving all existing correlations between variables stored in a Monte Carlo pre-computed phase space (PS) file, for dose calculation and high-resolution portal image prediction. The reference PS file was calculated using the PENELOPE code, after the flattening filter (FF) of an Elekta Synergy 6 MV photon beam. Each particle was represented in a mobile coordinate system by its radial position (r s ) in the PS plane, its energy (E), and its polar and azimuthal angles (φ d and θ d ), describing the particle deviation compared to its initial direction after bremsstrahlung, and the deviation orientation. Three sub-sources were created by sorting out particles according to their last interaction location (target, primary collimator or FF). For each sub-source, 4D correlated-histograms were built by storing E, r s , φ d and θ d values. Five different adaptive binning schemes were studied to construct 4D histograms of the VSMs, to ensure histogram efficient handling as well as an accurate reproduction of E, r s , φ d and θ d distribution details. The five resulting VSMs were then implemented in PENELOPE. Their accuracy was first assessed in the PS plane, by comparing E, r s , φ d and θ d distributions with those obtained from the reference PS file. Second, dose distributions computed in water, using the VSMs and the reference PS file located below the FF, and also after collimation in both water and heterogeneous phantom, were compared using a 1.5%–0 mm and a 2%–0 mm global gamma index, respectively. Finally, portal images were calculated without and with phantoms in the beam. The model was then evaluated using a 1%–0 mm global gamma index. Performance of a mono-source VSM was also investigated and led, as with the multi-source model, to excellent results when combined with an adaptive binning scheme.
Chabert, I; Barat, E; Dautremer, T; Montagu, T; Agelou, M; Croc de Suray, A; Garcia-Hernandez, J C; Gempp, S; Benkreira, M; de Carlan, L; Lazaro, D
2016-07-21
This work aims at developing a generic virtual source model (VSM) preserving all existing correlations between variables stored in a Monte Carlo pre-computed phase space (PS) file, for dose calculation and high-resolution portal image prediction. The reference PS file was calculated using the PENELOPE code, after the flattening filter (FF) of an Elekta Synergy 6 MV photon beam. Each particle was represented in a mobile coordinate system by its radial position (r s ) in the PS plane, its energy (E), and its polar and azimuthal angles (φ d and θ d ), describing the particle deviation compared to its initial direction after bremsstrahlung, and the deviation orientation. Three sub-sources were created by sorting out particles according to their last interaction location (target, primary collimator or FF). For each sub-source, 4D correlated-histograms were built by storing E, r s , φ d and θ d values. Five different adaptive binning schemes were studied to construct 4D histograms of the VSMs, to ensure histogram efficient handling as well as an accurate reproduction of E, r s , φ d and θ d distribution details. The five resulting VSMs were then implemented in PENELOPE. Their accuracy was first assessed in the PS plane, by comparing E, r s , φ d and θ d distributions with those obtained from the reference PS file. Second, dose distributions computed in water, using the VSMs and the reference PS file located below the FF, and also after collimation in both water and heterogeneous phantom, were compared using a 1.5%-0 mm and a 2%-0 mm global gamma index, respectively. Finally, portal images were calculated without and with phantoms in the beam. The model was then evaluated using a 1%-0 mm global gamma index. Performance of a mono-source VSM was also investigated and led, as with the multi-source model, to excellent results when combined with an adaptive binning scheme. PMID:27353090
The PENELOPE code system. Specific features and recent improvements
Highlights: • PENELOPE implements state-of-the-art models for electron and photon interactions. • It is characterized by a systematic use of class-II tracking of charged particles. • The code includes elaborate variance reduction methods and flexible geometry tools. - Abstract: Since its first release, back in 1996, the Monte Carlo code system PENELOPE has evolved into both a flexible and reliable tool for describing coupled electron–photon transport in complex material structures. The present article contains an overview of the physical interaction models, particle tracking methods, geometry tools, and variance-reduction techniques implemented in PENELOPE. Recent refinements aimed at improving the accuracy of the code, and its stability under variations of user-defined simulation parameters, are also described. These include the use of reliable cross sections for the ionization of inner atomic electron shells by electron/positron impact, a reformulation of the random-hinge method, and the use of fuzzy quadric surfaces in the description of the geometry
Vega Ramirez, J.L.; Chen, F.; Nicolucci, P.; Baffa, O. [Universidade de Sao Paulo (FFCLRP/USP), Ribeirao Preto, SP (Brazil). Faculdade de Filosofia, Ciencias e Letras. Dept. de Fisica e Matematica
2009-07-01
The dosimetric system of L-alanine mini dosimeter and K-Band EPR spectrometer was tested for the dosimetry in non-homogeneous media through the determination of the Percentage Depth Dose (PDD) curve for a small radiation field. The alanine mini dosimeters were produced by mechanical pressure of a mixture of L-alanine (95%) and PVA (5%) to nominal dimensions of 1 mm diameter and 3 mm length and 3 - 4 mg. For detecting the EPR signal of the mini dosimeters irradiated to 25 Gy, a K-Band (24 GHz) spectrometer was used. The dosimeters were irradiated in a {sup 60}Co radiotherapy unit using 80 cm source skin distance and field sizes of 2.5 x 2.5 cm{sup 2}. The inhomogeneous phantom consisted of acrylic and cork sheets of 30 x 30 x 1 cm{sup 3}; six cork sheets were sandwiched between five and nine acrylic sheets, which were placed at the top and bottom regions respectively. PDD curves with radiographic film and PENELOPE simulation were also determined. The PDD results for alanine mini dosimeters agreed better than 5.9% with film and PENELOPE. (author)
penORNL: a parallel Monte Carlo photon and electron transport package using PENELOPE
Bekar, Kursat B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, Thomas Martin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Weber, Charles F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2015-01-01
The parallel Monte Carlo photon and electron transport code package penORNL was developed at Oak Ridge National Laboratory to enable advanced scanning electron microscope (SEM) simulations on high performance computing systems. This paper discusses the implementations, capabilities and parallel performance of the new code package. penORNL uses PENELOPE for its physics calculations and provides all available PENELOPE features to the users, as well as some new features including source definitions specifically developed for SEM simulations, a pulse-height tally capability for detailed simulations of gamma and x-ray detectors, and a modified interaction forcing mechanism to enable accurate energy deposition calculations. The parallel performance of penORNL was extensively tested with several model problems, and very good linear parallel scaling was observed with up to 512 processors. penORNL, along with its new features, will be available for SEM simulations upon completion of the new pulse-height tally implementation.
A general vision of EGS4 and PENELOPE codes and perspectives of utilization in radiotherapy
In Brazil, the lack of works in Radiotherapy simulations is huge and the perspectives are good. The works developed now in some research centers show the viability and the crescent interest of applications in this area. In this work a general vision of Monte Carlo method application is presented. A comparison of EGS4 and PENELOPE codes show their differences in applications and perspectives in Radiotherapy. (author)
PeneloPET, a Monte Carlo PET simulation tool based on PENELOPE: features and validation
Espana, S; Herraiz, J L; Vicente, E; Udias, J M [Grupo de Fisica Nuclear, Departmento de Fisica Atomica, Molecular y Nuclear, Universidad Complutense de Madrid, Madrid (Spain); Vaquero, J J; Desco, M [Unidad de Medicina y CirugIa Experimental, Hospital General Universitario Gregorio Maranon, Madrid (Spain)], E-mail: jose@nuc2.fis.ucm.es
2009-03-21
Monte Carlo simulations play an important role in positron emission tomography (PET) imaging, as an essential tool for the research and development of new scanners and for advanced image reconstruction. PeneloPET, a PET-dedicated Monte Carlo tool, is presented and validated in this work. PeneloPET is based on PENELOPE, a Monte Carlo code for the simulation of the transport in matter of electrons, positrons and photons, with energies from a few hundred eV to 1 GeV. PENELOPE is robust, fast and very accurate, but it may be unfriendly to people not acquainted with the FORTRAN programming language. PeneloPET is an easy-to-use application which allows comprehensive simulations of PET systems within PENELOPE. Complex and realistic simulations can be set by modifying a few simple input text files. Different levels of output data are available for analysis, from sinogram and lines-of-response (LORs) histogramming to fully detailed list mode. These data can be further exploited with the preferred programming language, including ROOT. PeneloPET simulates PET systems based on crystal array blocks coupled to photodetectors and allows the user to define radioactive sources, detectors, shielding and other parts of the scanner. The acquisition chain is simulated in high level detail; for instance, the electronic processing can include pile-up rejection mechanisms and time stamping of events, if desired. This paper describes PeneloPET and shows the results of extensive validations and comparisons of simulations against real measurements from commercial acquisition systems. PeneloPET is being extensively employed to improve the image quality of commercial PET systems and for the development of new ones.
An investigation on the capabilities of the PENELOPE MC code in nanodosimetry
The Monte Carlo (MC) method has been widely implemented in studies of radiation effects on human genetic material. Most of these works have used specific-purpose MC codes to simulate radiation transport in condensed media. PENELOPE is one of the general-purpose MC codes that has been used in many applications related to radiation dosimetry. Based on the fact that PENELOPE can carry out event-by-event coupled electron-photon transport simulations following these particles down to energies of the order of few tens of eV, we have decided to investigate the capacities of this code in the field of nanodosimetry. Single and double strand break probabilities due to the direct impact of γ rays originated from Co60 and Cs137 isotopes and characteristic x-rays, from Al and C K-shells, have been determined by use of PENELOPE. Indirect damage has not been accounted for in this study. A human genetic material geometrical model has been developed, taking into account five organizational levels. In an article by Friedland et al. [Radiat. Environ. Biophys. 38, 39-47 (1999)], a specific-purpose MC code and a very sophisticated DNA geometrical model were used. We have chosen that work as a reference to compare our results. Single and double strand-break probabilities obtained here underestimate those reported by Friedland and co-workers by 20%-76% and 50%-60%, respectively. However, we obtain RBE values for Cs137, AlK and CK radiations in agreement with those reported in previous works [Radiat. Environ. Biophys. 38, 39-47 (1999)] and [Phys. Med. Biol. 53, 233-244 (2008)]. Some enhancements can be incorporated into the PENELOPE code to improve its results in the nanodosimetry field.
An investigation on the capabilities of the PENELOPE MC code in nanodosimetry
Bernal, M. A.; Liendo, J. A. [Departamento de Fisica, Universidad Simon Bolivar, P.O. Box 89000, Caracas (Venezuela, Bolivarian Republic of)
2009-02-15
The Monte Carlo (MC) method has been widely implemented in studies of radiation effects on human genetic material. Most of these works have used specific-purpose MC codes to simulate radiation transport in condensed media. PENELOPE is one of the general-purpose MC codes that has been used in many applications related to radiation dosimetry. Based on the fact that PENELOPE can carry out event-by-event coupled electron-photon transport simulations following these particles down to energies of the order of few tens of eV, we have decided to investigate the capacities of this code in the field of nanodosimetry. Single and double strand break probabilities due to the direct impact of {gamma} rays originated from Co{sup 60} and Cs{sup 137} isotopes and characteristic x-rays, from Al and C K-shells, have been determined by use of PENELOPE. Indirect damage has not been accounted for in this study. A human genetic material geometrical model has been developed, taking into account five organizational levels. In an article by Friedland et al. [Radiat. Environ. Biophys. 38, 39-47 (1999)], a specific-purpose MC code and a very sophisticated DNA geometrical model were used. We have chosen that work as a reference to compare our results. Single and double strand-break probabilities obtained here underestimate those reported by Friedland and co-workers by 20%-76% and 50%-60%, respectively. However, we obtain RBE values for Cs{sup 137}, Al{sub K} and C{sub K} radiations in agreement with those reported in previous works [Radiat. Environ. Biophys. 38, 39-47 (1999)] and [Phys. Med. Biol. 53, 233-244 (2008)]. Some enhancements can be incorporated into the PENELOPE code to improve its results in the nanodosimetry field.
An investigation on the capabilities of the PENELOPE MC code in nanodosimetry.
Bernal, M A; Liendo, J A
2009-02-01
The Monte Carlo (MC) method has been widely implemented in studies of radiation effects on human genetic material. Most of these works have used specific-purpose MC codes to simulate radiation transport in condensed media. PENELOPE is one of the general-purpose MC codes that has been used in many applications related to radiation dosimetry. Based on the fact that PENELOPE can carry out event-by-event coupled electron-photon transport simulations following these particles down to energies of the order of few tens of eV, we have decided to investigate the capacities of this code in the field of nanodosimetry. Single and double strand break probabilities due to the direct impact of gamma rays originated from Co60 and Cs137 isotopes and characteristic x-rays, from Al and C K-shells, have been determined by use of PENELOPE. Indirect damage has not been accounted for in this study. A human genetic material geometrical model has been developed, taking into account five organizational levels. In an article by Friedland et al. [Radiat. Environ. Biophys. 38, 39-47 (1999)], a specific-purpose MC code and a very sophisticated DNA geometrical model were used. We have chosen that work as a reference to compare our results. Single and double strand-break probabilities obtained here underestimate those reported by Friedland and co-workers by 20%-76% and 50%-60%, respectively. However, we obtain RBE values for Cs137, AlK and CK radiations in agreement with those reported in previous works [Radiat. Environ. Biophys. 38, 39-47 (1999)] and [Phys. Med. Biol. 53, 233-244 (2008)]. Some enhancements can be incorporated into the PENELOPE code to improve its results in the nanodosimetry field. PMID:19292002
We present Monte Carlo simulations of the gamma exposure in closed rooms made of steel or concrete and contaminated by 60Co or NORM radionuclides. The computer code PENELOPE-2008 (Salvat et al., 2009) was used. Our simulations for 60Co suggest considering detailed Monte Carlo simulations in future recommendations on clearance and exemption of materials with low radioactivity. For NORM nuclides our calculations suggest that Monte Carlo simulations are a possible alternative in case a material fails the dose rate criteria by using the RP 112 screening method. - Highlights: • PENELOPE-2008 was used for Monte Carlo simulations of gamma exposure in closed rooms made of steel or concrete. • Findings support introducing IAEA SR 44 activity concentration value of 0.1 Bq/g as exemption value for 60Co. • PENELOPE-2008 calculations show good agreement with a density corrected Berger model for dose rate calculations concerning NORM building materials. • Monte Carlo calculations or a density corrected Berger model could be used to modify the model suggested in RP 112
Almansa, J F; Anguiano, M; Guerrero, R; Lallena, A M; Al-Dweri, Feras M.O.; Almansa, Julio F.; Guerrero, Rafael
2006-01-01
Monte Carlo calculations using the codes PENELOPE and GEANT4 have been performed to characterize the dosimetric properties of monoenergetic photon point sources in water. The dose rate in water has been calculated for energies of interest in brachytherapy, ranging between 10 keV and 2 MeV. A comparison of the results obtained using the two codes with the available data calculated with other Monte Carlo codes is carried out. A chi2-like statistical test is proposed for these comparisons. PENELOPE and GEANT4 show a reasonable agreement for all energies analyzed and distances to the source larger than 1 cm. Significant differences are found at distances from the source up to 1 cm. A similar situation occurs between PENELOPE and EGS4.
Monte Carlo Simulation of Secondary Fluorescence using a New Graphical Interface for PENELOPE
Pinard, P. T.; Demers, H.; Llovet, X.; Gauvin, R.; Salvat, F.
2011-12-01
Secondary fluorescence is not a negligible factor in the chemical concentration measurement of many minerals (quartz, olivine, etc.) using the electron probe microanalysis (EPMA) technique (Llovet and Galán, 2003). The importance of this phenomenon depends on the chemical species present in the mineral but also, in case of heterogeneous samples, on their relative location to the measurement position. Monte Carlo codes are useful tools to select the optimal measurement conditions as well as to correct afterwards the results for phenomenon such as secondary fluorescence. PENELOPE (Salvat et al., 2011) is a Fortran Monte Carlo code for simulation of coupled electron-photon transport in matter that allows a detailed interpretation of experimental results of electron spectroscopy and microscopy. PENEPMA is a dedicated main program of PENELOPE designed to perform simulations with the same parameters as in actual EPMA measurements. Complex geometries can be defined to emulate the internal structure of a sample. Photon interactions are simulated in chronological succession, therefore allowing the calculation of secondary fluorescence. These features combined with the use of the most reliable physical interaction models make PENEPMA a unique Monte Carlo code for EPMA analysis. However, the original version of PENEPMA had a steep learning curve as it required the user to manually create several input files to run a single simulation. To facilitate the use of the code, a graphical interface was recently developed. Written in the cross-platform programming language Python, it simplifies the setup of simulations and the analysis of the results. It also includes optimized simulation parameters which increases the efficiency of the simulations (i.e. reduces the computation time) by a factor of up to 8. In this communication, we describe the structure and capabilities of this graphical interface. It not only eases the definition of the problem, but also provides more extensive
Vega R, J. L.; Nicolucci, P.; Baffa, O. [Universidade de Sao Paulo, FFCLRP, Departamento de Fisica, Av. Bandeirantes 3900, Bairro Monte Alegre, 14040-901 Ribeirao Preto, Sao Paulo (Brazil); Chen, F. [Universidade Federale do ABC, CCNH, Rua Santa Adelia 166, Bangu, 09210-170 Santo Andre, Sao Paulo (Brazil); Apaza V, D. G., E-mail: josevegaramirez@yahoo.es [Universidad Nacional de San Agustin de Arequipa, Departamento de Fisica, Arequipa (Peru)
2014-08-15
The dosimetry system based on alanine mini dosimeters plus K-Band EPR spectrometer was tested in the tissue-interface dosimetry through the percentage depth-dose (Pdd) determination for 3 x 3 cm{sup 2} and 1 x 1 cm{sup 2} radiation fields sizes. The alanine mini dosimeters were produced by mechanical pressure from a mixture of 95% L-alanine and 5% polyvinyl alcohol (Pva) acting as binder. Nominal dimensions of these mini dosimeters were 1 mm diameter and 3 mm length as well as 3 - 4 mg mass. The EPR spectra of the mini dosimeters were registered using a K-Band (24 GHz) EPR spectrometer. The mini dosimeters were placed in a nonhomogeneous phantom and irradiated with 20 Gy in a 6 MV PRIMUS Siemens linear accelerator, with a source-to-surface distance of 100 cm using the small fields previously mentioned. The cylindrical non-homogeneous phantom was comprised of several disk-shaped plates of different materials in the sequence acrylic-bone cork-bone-acrylic, with dimensions 15 cm diameter and 1 cm thick. The plates were placed in descending order, starting from top with four acrylic plates followed by two bone plates plus eight cork plates plus two bone plates and finally, four acrylic plates (4-2-8-2-4). Pdd curves from the treatment planning system and from Monte Carlo simulation with Penelope code were determined. Mini dosimeters Pdd results show good agreement with Penelope, better than 95% for the cork homogeneous region and 97.7% in the bone heterogeneous region. In the first interface region, between acrylic and bone, it can see a dose increment of 0.6% for mini dosimeters compared to Penelope. At the second interface, between bone and cork, there is 9.1% of dose increment for mini dosimeter relative to Penelope. For the third (cork-bone) and fourth (bone-acrylic) interfaces, the dose increment for mini dosimeters compared to Penelope was 4.1% both. (Author)
Rojas C, E. L.; Avila, O., E-mail: leticia.rojas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)
2012-10-15
In this work the simulation codes Monte Carlo, Penelope and MCNPX were used to calculate the doses by unit of accumulated activity S(N-N) in water spherical cells models of different radius exposed to mono-energetics electrons coming from punctual sources located in the center of the cellular nucleus. The studied cellular radii were: r{sub n}1=3 r{sub c}1=6; r{sub n}2=5 and r{sub c}2=10; r{sub n}3=9 and r{sub c}3=10 {mu}m; being r{sub n} and r{sub c} the nuclear and cellular radius, respectively. The following initial energies of the electrons were considered: 1, 5, 10, 50, 100, 500, 700 and 1000 keV. Additionally values S(N-N) were calculated for spherical cells of r= 3 {mu}m r{sub c}= 6 {mu}m due to the electrons coming from sources of {sup 111}In, {sup 177}Lu, {sup 99m}Tc, {sup 188}Re and {sup 186}Re. The obtained values are compared with those calculated by the MIRD Committee internationally accepted. The percentage differences between the values reported by this Committee and those calculated by Monte Carlo simulation are inside the interval that is considered valid for this dosimetry type. A major concordance was found among the values calculated by Monte Carlo simulation that among those calculated by MIRD and those obtained by simulation. Considering validated the use of both codes for similar applications, the values S(N-N) and S(N-C y) were obtained of prostate cancer real cells models of the PC3 line. The results were compared among them. The values of S(N-N) obtained with Penelope for the PC3 cells for the electron emissions of {sup 111}In, {sup 177}Lu, {sup 99m}Tc, {sup 188}Re and {sup 186}Re are: 3.19e{sup {sub {sup 4}}}, 3.24e{sup -4}, 1.37e{sup -4}, 1.11e{sup -4} and 1.91e{sup -4} Gy/Bq-s, respectively. Also the obtained results for S(N-C y) are: 2.95e{sup -6}, 3.17e{sup -5}, 2.09e{sup -6}, 1.41e{sup -5}, 1.86e{sup -5} Gy/Bq-s. (Author)
A new comparison of the values published for the chamber quality factor fc,Q0 of a NACP-02 plane-parallel chamber in 60Co, calculated with the Monte Carlo (MC) systems EGSnrc and PENELOPE, shows a difference of approximately 0.5%. The authors analyse possible reasons for this difference and recalculate the chamber quality factor with EGSnrc. Variations in the simulation transport parameters of EGSnrc result in changes smaller than the difference. An investigation of the most important uncertainties of cross-sectional data, considering variations in the mean excitation energy of stopping powers and in the total photon cross-sections, shows uncertainties comparable to the differences between the two codes for the chamber quality factor. Variations of the front wall thickness of the NACP-02 chamber, in the range of discrepancies with manufacturer data reported in the literature, result in significant changes in the calculated values. However, the difference in fc,Q0 cannot be explained in terms of these modifications. Hence, although both codes have been demonstrated to yield artefact free ion chamber simulations, a convergence of results for this particular problem cannot be achieved. An uncertainty estimate which takes into account the 0.5% difference for the MC calculated chamber quality factors seems to be a reasonable assumption. (author)
Rodriguez, Miguel; Sempau, Josep [Institut de Tècniques Energètiques, Universitat Politècnica de Catalunya, Diagonal 647, Barcelona E-08028 (Spain); Brualla, Lorenzo, E-mail: lorenzo.brualla@uni-duisburg-essen.de [NCTeam, Strahlenklinik, Universitätsklinikum Essen, Hufelandstraße 55, Essen D-45122 (Germany)
2015-06-15
Purpose: The Monte Carlo simulation of electron transport in Linac targets using the condensed history technique is known to be problematic owing to a potential dependence of absorbed dose distributions on the electron step length. In the PENELOPE code, the step length is partially determined by the transport parameters C1 and C2. The authors have investigated the effect on the absorbed dose distribution of the values given to these parameters in the target. Methods: A monoenergetic 6.26 MeV electron pencil beam from a point source was simulated impinging normally on a cylindrical tungsten target. Electrons leaving the tungsten were discarded. Radial absorbed dose profiles were obtained at 1.5 cm of depth in a water phantom located at 100 cm for values of C1 and C2 in the target both equal to 0.1, 0.01, or 0.001. A detailed simulation case was also considered and taken as the reference. Additionally, lateral dose profiles were estimated and compared with experimental measurements for a 6 MV photon beam of a Varian Clinac 2100 for the cases of C1 and C2 both set to 0.1 or 0.001 in the target. Results: On the central axis, the dose obtained for the case C1 = C2 = 0.1 shows a deviation of (17.2% ± 1.2%) with respect to the detailed simulation. This difference decreases to (3.7% ± 1.2%) for the case C1 = C2 = 0.01. The case C1 = C2 = 0.001 produces a radial dose profile that is equivalent to that of the detailed simulation within the reached statistical uncertainty of 1%. The effect is also appreciable in the crossline dose profiles estimated for the realistic geometry of the Linac. In another simulation, it was shown that the error made by choosing inappropriate transport parameters can be masked by tuning the energy and focal spot size of the initial beam. Conclusions: The use of large path lengths for the condensed simulation of electrons in a Linac target with PENELOPE conducts to deviations of the dose in the patient or phantom. Based on the results obtained in
Habib, B.; Poumarede, B.; Tola, F.; Barthe, J. [CEA, LIST, Dept Technol Capteur et Signal, F-91191 Gif Sur Yvette, (France)
2010-07-01
The aim of the present study is to demonstrate the potential of accelerated dose calculations, using the fast Monte Carlo (MC) code referred to as PENFAST, rather than the conventional MC code PENELOPE, without losing accuracy in the computed dose. For this purpose, experimental measurements of dose distributions in homogeneous and inhomogeneous phantoms were compared with simulated results using both PENELOPE and PENFAST. The simulations and experiments were performed using a Saturne 43 linac operated at 12 MV (photons), and at 18 MeV (electrons). Pre-calculated phase space files (PSFs) were used as input data to both the PENELOPE and PENFAST dose simulations. Since depth-dose and dose profile comparisons between simulations and measurements in water were found to be in good agreement (within {+-} 1% to 1 mm), the PSF calculation is considered to have been validated. In addition, measured dose distributions were compared to simulated results in a set of clinically relevant, inhomogeneous phantoms, consisting of lung and bone heterogeneities in a water tank. In general, the PENFAST results agree to within a 1% to 1 mm difference with those produced by PENELOPE, and to within a 2% to 2 mm difference with measured values. Our study thus provides a pre-clinical validation of the PENFAST code. It also demonstrates that PENFAST provides accurate results for both photon and electron beams, equivalent to those obtained with PENELOPE. CPU time comparisons between both MC codes show that PENFAST is generally about 9-21 times faster than PENELOPE. (authors)
Comparative Dosimetric Estimates of a 25 keV Electron Micro-beam with three Monte Carlo Codes
The calculations presented compare the different performances of the three Monte Carlo codes PENELOPE-1999, MCNP-4C and PITS, for the evaluation of Dose profiles from a 25 keV electron micro-beam traversing individual cells. The overall model of a cell is a water cylinder equivalent for the three codes but with a different internal scoring geometry: hollow cylinders for PENELOPE and MCNP, whereas spheres are used for the PITS code. A cylindrical cell geometry with scoring volumes with the shape of hollow cylinders was initially selected for PENELOPE and MCNP because of its superior simulation of the actual shape and dimensions of a cell and for its improved computer-time efficiency if compared to spherical internal volumes. Some of the transfer points and energy transfer that constitute a radiation track may actually fall in the space between spheres, that would be outside the spherical scoring volume. This internal geometry, along with the PENELOPE algorithm, drastically reduced the computer time when using this code if comparing with event-by-event Monte Carlo codes like PITS. This preliminary work has been important to address dosimetric estimates at low electron energies. It demonstrates that codes like PENELOPE can be used for Dose evaluation, even with such small geometries and energies involved, which are far below the normal use for which the code was created. Further work (initiated in Summer 2002) is still needed however, to create a user-code for PENELOPE that allows uniform comparison of exact cell geometries, integral volumes and also microdosimetric scoring quantities, a field where track-structure codes like PITS, written for this purpose, are believed to be superior
Comparative Dosimetric Estimates of a 25 keV Electron Micro-beam with three Monte Carlo Codes
Mainardi, Enrico; Donahue, Richard J.; Blakely, Eleanor A.
2002-09-11
The calculations presented compare the different performances of the three Monte Carlo codes PENELOPE-1999, MCNP-4C and PITS, for the evaluation of Dose profiles from a 25 keV electron micro-beam traversing individual cells. The overall model of a cell is a water cylinder equivalent for the three codes but with a different internal scoring geometry: hollow cylinders for PENELOPE and MCNP, whereas spheres are used for the PITS code. A cylindrical cell geometry with scoring volumes with the shape of hollow cylinders was initially selected for PENELOPE and MCNP because of its superior simulation of the actual shape and dimensions of a cell and for its improved computer-time efficiency if compared to spherical internal volumes. Some of the transfer points and energy transfer that constitute a radiation track may actually fall in the space between spheres, that would be outside the spherical scoring volume. This internal geometry, along with the PENELOPE algorithm, drastically reduced the computer time when using this code if comparing with event-by-event Monte Carlo codes like PITS. This preliminary work has been important to address dosimetric estimates at low electron energies. It demonstrates that codes like PENELOPE can be used for Dose evaluation, even with such small geometries and energies involved, which are far below the normal use for which the code was created. Further work (initiated in Summer 2002) is still needed however, to create a user-code for PENELOPE that allows uniform comparison of exact cell geometries, integral volumes and also microdosimetric scoring quantities, a field where track-structure codes like PITS, written for this purpose, are believed to be superior.
The progress in cancer treatment systems in heterogeneities of human body has had obstacles by the lack of a suitable experimental model test. The only option is to develop simulated theoretical models that have the same properties in interfaces similar to human tissues, to know the radiation behavior in the interaction with these materials. In this paper we used the Monte Carlo method by Penelope code based solely on studies for the cancer treatment as well as for the calibration of beams and their various interactions in mannequins. This paper also aims the construction, simulation and characterization of an equivalent object to the tissues of the human body with various heterogeneities, we will later use to control and plan experientially doses supplied in treating tumors in radiotherapy. To fulfill the objective we study the ionizing radiation and the various processes occurring in the interaction with matter; understanding that to calculate the dose deposited in tissues interfaces (percentage depth dose) must be taken into consideration aspects such as the deposited energy, irradiation fields, density, thickness, tissue sensitivity and other items. (Author)
Guaspari, David
1995-01-01
A formal program verification is a (mathematical) proof that a program executed according to its intended model meets some specification. This proves that the algorithm defined by the program is correct in the precise technical sense of being consistent with a particular specification. A program correct in this sense is free from a large and important class of errors, even though its behavior may still produce unintended results--either because the implementation of the programming language itself does not match the model of execution, or because the specification does not correctly express the user's intentions. Penelope is a prototype system for interactively developing and verifying programs that are written in a rich subset of sequential Ada. Penelope can be used to develop a program and its correctness proof incrementally, and in concert with one another. Incrementality is used in a number of ways to help make verification more tractable and more productive. For example, if an already-verified program is modified, one can attempt to prove the modified version by replaying and modifying the original verification. Penelope's specification language, Larch/Ada, belongs to the family of Larch interface languages. Larch/Ada scales up properly, in the sense that it is demonstrably sound to decompose a system hierarchically and reason locally about the implementation of each piece. Penelope has been applied in various demonstration projects--for specification (guidance control, distributed operating systems), verification (of off-the-shelf code), and formal development (by non-expert as well as expert users). Some features of Penelope have been embodied in Ada Wise, a lint-like non-interactive tool that warns of the potential for certain dynamic semantic errors in Ada programs.
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.)
Penelope simulations of photon calibration fields at the LMIV-IPEN, Brazil
Kakoi, Adelia Aparecida Yuka; Heredia, Eduardo; Xavier, Marcos; Rodrigues Junior, Orlando; Cardoso, Joaquim C.S., E-mail: adelia@usp.b, E-mail: jcardoso@ipen.b, E-mail: rodrijr@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2011-07-01
In this article, the photon spectra of Eu-152 reference photon field of the detection system at LMIV-IPEN is presented, which was calculated by means of Monte Carlo simulations by using the computer transport code Penelope. The contributions from scattered photons to the spectra of the fields have been determined with regard to the quantities photon fluence. The mean photon energies calculated with respect to the mentioned quantity are listed. Differences in the design of the sources and their influence on the spectra are discussed. The dependence of the scattered photon component from the energy was examined by feeding the Penelope code with monoenergetic photons of different energies. (author)
PENELOPE. Taking advantage of Class II simulation
PENELOPE is a general-purpose Monte Carlo code for the simulation of electron-photon showers in arbitrary materials. It covers the energy range from ∼1 keV up to ∼1 GeV. The simulation of electron and positron transport is based on a mixed (Class II) simulation scheme. The cut-off angular deflection, which separates soft and hard (catastrophic) elastic collisions, is allowed to vary with the energy of the particle in such a way that the multiple soft interactions which occur between two consecutive hard events produce only gentle deflections of the track. Space displacements are generated by a simple and accurate algorithm that works even in the vicinity of interfaces. The physics of PENELOPE, the simulation algorithm for charged particle transport and the structure and operation of the code system are described. The distribution package also includes subroutines for simulation in quadric geometries (i.e. material systems consisting of homogeneous bodies limited by quadric surfaces) and a simple geometry viewer. (author)
A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom
Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H., E-mail: mbellezzo@gmail.br [Instituto de Pesquisas Energeticas e Nucleares / CNEN, Av. Lineu Prestes 2242, Cidade Universitaria, 05508-000 Sao Paulo (Brazil)
2014-08-15
As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)
A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom
As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)
Cardena R, A. R.; Vega R, J. L.; Apaza V, D. G., E-mail: cardroj@yahoo.es [Universidad Nacional de San Agustin, Av. Independencia s/n, Arequipa (Peru)
2015-10-15
The progress in cancer treatment systems in heterogeneities of human body has had obstacles by the lack of a suitable experimental model test. The only option is to develop simulated theoretical models that have the same properties in interfaces similar to human tissues, to know the radiation behavior in the interaction with these materials. In this paper we used the Monte Carlo method by Penelope code based solely on studies for the cancer treatment as well as for the calibration of beams and their various interactions in mannequins. This paper also aims the construction, simulation and characterization of an equivalent object to the tissues of the human body with various heterogeneities, we will later use to control and plan experientially doses supplied in treating tumors in radiotherapy. To fulfill the objective we study the ionizing radiation and the various processes occurring in the interaction with matter; understanding that to calculate the dose deposited in tissues interfaces (percentage depth dose) must be taken into consideration aspects such as the deposited energy, irradiation fields, density, thickness, tissue sensitivity and other items. (Author)
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)
SPQR: a Monte Carlo reactor kinetics code
The SPQR Monte Carlo code has been developed to analyze fast reactor core accident problems where conventional methods are considered inadequate. The code is based on the adiabatic approximation of the quasi-static method. This initial version contains no automatic material motion or feedback. An existing Monte Carlo code is used to calculate the shape functions and the integral quantities needed in the kinetics module. Several sample problems have been devised and analyzed. Due to the large statistical uncertainty associated with the calculation of reactivity in accident simulations, the results, especially at later times, differ greatly from deterministic methods. It was also found that in large uncoupled systems, the Monte Carlo method has difficulty in handling asymmetric perturbations
Coded aperture optimization using Monte Carlo simulations
Coded apertures using Uniformly Redundant Arrays (URA) have been unsuccessfully evaluated for two-dimensional and three-dimensional imaging in Nuclear Medicine. The images reconstructed from coded projections contain artifacts and suffer from poor spatial resolution in the longitudinal direction. We introduce a Maximum-Likelihood Expectation-Maximization (MLEM) algorithm for three-dimensional coded aperture imaging which uses a projection matrix calculated by Monte Carlo simulations. The aim of the algorithm is to reduce artifacts and improve the three-dimensional spatial resolution in the reconstructed images. Firstly, we present the validation of GATE (Geant4 Application for Emission Tomography) for Monte Carlo simulations of a coded mask installed on a clinical gamma camera. The coded mask modelling was validated by comparison between experimental and simulated data in terms of energy spectra, sensitivity and spatial resolution. In the second part of the study, we use the validated model to calculate the projection matrix with Monte Carlo simulations. A three-dimensional thyroid phantom study was performed to compare the performance of the three-dimensional MLEM reconstruction with conventional correlation method. The results indicate that the artifacts are reduced and three-dimensional spatial resolution is improved with the Monte Carlo-based MLEM reconstruction.
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
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...
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)
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)
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)
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
Vilches, M.; García-Pareja, S.; Guerrero, R.; Anguiano, M.; Lallena, A. M.
2007-09-01
When a therapeutic electron linear accelerator is simulated using a Monte Carlo (MC) code, the tuning of the initial spectra and the renormalization of dose (e.g., to maximum axial dose) constitute a common practice. As a result, very similar depth dose curves are obtained for different MC codes. However, if renormalization is turned off, the results obtained with the various codes disagree noticeably. The aim of this work is to investigate in detail the reasons of this disagreement. We have found that the observed differences are due to non-negligible differences in the angular scattering of the electron beam in very thin slabs of dense material (primary foil) and thick slabs of very low density material (air). To gain insight, the effects of the angular scattering models considered in various MC codes on the dose distribution in a water phantom are discussed using very simple geometrical configurations for the LINAC. The MC codes PENELOPE 2003, PENELOPE 2005, GEANT4, GEANT3, EGSnrc and MCNPX have been used.
FLUKA and PENELOPE simulations of 10 keV to 10 MeV photons in LYSO and soft tissue
Monte Carlo simulations of electromagnetic particle interactions and transport by FLUKA and PENELOPE were compared. 10 keV to 10 MeV incident photon beams impinged a LYSO crystal and a soft-tissue phantom. Central-axis as well as off-axis depth doses agreed within 1 s.d.; no systematic under- or over-estimate of the pulse height spectra was observed from 100 keV to 10 MeV for both materials, agreement was within 5%. Simulation of photon and electron transport and interactions at this level of precision and reliability is of significant impact, for instance, on treatment monitoring of hadrontherapy where a code like FLUKA is needed to simulate the full suite of particles and interactions (not just electromagnetic). At the interaction-by-interaction level, apart from known differences in condensed history techniques, two-quanta positron annihilation at rest was found to differ between the two codes. PENELOPE produced a 511 keV sharp line, whereas FLUKA produced visible acolinearity, a feature recently implemented to account for the momentum of shell electrons. - Highlights: • Monte Carlo simulations of electromagnetic particle interactions and transport by FLUKA and PENELOPE were compared. • 10 keV to 10 MeV incident photon beams impinged a LYSO crystal and a soft-tissue phantom. • The pulse height spectra, depth doses central-axis as well as off-axis were found to agree within statistical uncertainty; no systematic difference was observed
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)
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)
Photon fluence spectra of the Seibersdorf Labor/BEV Picker 60Co therapy unit were calculated using two generally recognised Monte Carlo codes, PENELOPE-2006 and MCNP5. The complexity of the simulation model was increased in three steps (from a pure source capsule and a simplified model using rotational symmetry to a realistic model of the facility). Photon fluence spectra of both codes generally agree within their statistical standard uncertainties for the case of identical geometry set-up and particle transport parameter settings. Resulting total fluence values were about 0.3% higher for MCNP as compared to PENELOPE. The verification of the simulated photon fluence spectra was based upon depth-dose measurements in water performed with a PTW 31003 ionisation chamber and a thick-walled chamber type CC01. The depth-dose curve calculated with PENELOPE agreed with the curve obtained from measurements within 0.4% across the available depth region in the 30 cm x 30 cm x 30 cm water phantom. The comparison of measured and simulated beam quality indices (TPR20,10) revealed deviations of less than 0.2%.
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)
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
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)
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...
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...
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
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)
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
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
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)
Al-Dweri, F M O; Rojas, E L; Al-Dweri, Feras M.O.; Lallena, Antonio M.
2005-01-01
Monte Carlo simulation with PENELOPE (v.~2003) is applied to calculate Leksell Gamma Knife$^{\\circledR}$ dose distributions for heterogeneous phantoms. The usual spherical water phantom is modified with a spherical bone shell simulating the skull and an air-filled cube simulating the frontal or maxillary sinuses. Different simulations of the 201 source configuration of the Gamma Knife have been carried out with a simplified model of the geometry of the source channel of the Gamma Knife recently tested for both single source and multisource configurations. The dose distributions determined for heterogeneous phantoms including the bone- and/or air-tissue interfaces show non negligible differences with respect to those calculated for a homogeneous one, mainly when the Gamma Knife isocenter approaches the separation surfaces. Our findings confirm an important underdosage ($\\sim$10%) nearby the air-tissue interface, in accordance with previous results obtained with PENELOPE code with a procedure different to ours....
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.
A method based on a combination of the variance-reduction techniques of particle splitting and Russian roulette is presented. This method improves the efficiency of radiation transport through linear accelerator geometries simulated with the Monte Carlo method. The method named as ‘splitting-roulette’ was implemented on the Monte Carlo code PENELOPE and tested on an Elekta linac, although it is general enough to be implemented on any other general-purpose Monte Carlo radiation transport code and linac geometry. Splitting-roulette uses any of the following two modes of splitting: simple splitting and ‘selective splitting’. Selective splitting is a new splitting mode based on the angular distribution of bremsstrahlung photons implemented in the Monte Carlo code PENELOPE. Splitting-roulette improves the simulation efficiency of an Elekta SL25 linac by a factor of 45. (paper)
Camila Salata
2009-08-01
Full Text Available OBJETIVO: Utilizar o código PENELOPE e desenvolver geometrias onde estão presentes heterogeneidades para simular o comportamento do feixe de fótons nessas condições. MATERIAIS E MÉTODOS: Foram feitas simulações do comportamento da radiação ionizante para o caso homogêneo, apenas água, e para os casos heterogêneos, com diferentes materiais. Consideraram-se geometrias cúbicas para os fantomas e geometrias em forma de paralelepípedos para as heterogeneidades com a seguinte composição: tecido simulador de osso e pulmão, seguindo recomendações da International Commission on Radiological Protection, e titânio, alumínio e prata. Definiram-se, como parâmetros de entrada: a energia e o tipo de partícula da fonte, 6 MV de fótons; a distância fonte-superfície de 100 cm; e o campo de radiação de 10x 10 cm². RESULTADOS: Obtiveram-se curvas de percentual de dose em profundidade para todos os casos. Observou-se que em materiais com densidade eletrônica alta, como a prata, a dose absorvida é maior em relação à dose absorvida no fantoma homogêneo, enquanto no tecido simulador de pulmão a dose é menor. CONCLUSÃO: Os resultados obtidos demonstram a importância de se considerar heterogeneidades nos algoritmos dos sistemas de planejamento usados no cálculo da distribuição de dose nos pacientes, evitando-se sub ou superdosagem dos tecidos próximos às heterogeneidades.OBJECTIVE: The PENELOPE code was utilized to simulate irradiation geometries where heterogeneities are present and to simulate a photon beam behavior under these conditions. MATERIALS AND METHODS: For the homogeneous case, the ionizing radiation behavior was simulated only with water, and different materials were introduced to simulate heterogeneous conditions. Cubic geometries were utilized for the homogeneous phantoms, and parallelepiped-shaped geometries for the heterogeneities with the following composition: bone and lung tissue simulators, as recommended
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)
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.
Lezin, G T; Makarova, K V; Velikodvorskaia, V V; Zelentsova, E S; Kechumian, R R; Kidwell, M G; Kunin, E V; Evgen'ev, M B
2001-01-01
The mobile element Penelope is activated and mobilizes several other transposons in dysgenic crosses in Drosophila virilis. Its structure proved to be complex and to vary greatly in all examined species of the virilis group. Phylogenetic analysis of the reverse transcriptase (RT) domain assigned Penelope to a new branch, rather than to any known family, of LTR-lacking retroelements. Amino acid sequence analysis showed that the C-terminal domain of the Penelope polyprotein is an active endonuclease, which is related to intron-encoded endonucleases and to bacterial repair endonuclease UrvC, and may act as an integras. Retroelements coding for a putative endonuclease that differs from typical integrase have thus far not been known. The N-terminal domain of the Penelope polyprotein was shown to contain a protease with significant homology to HIV-1 protease. Phylogenetic analysis divided the Penelope copies from several virilis species into two subfamilies, one including virtually identical full-length copies, and the other comprising highly divergent defective copies. The results suggest both vertical and horizontal transfer of the element. Possibly, Penelope invasion recurred during evolution and contributed to genome rearrangement in the virilis species. Chromosome aberrations detected in D. virilis, which is now being invaded by Penelope, is direct evidence for this assumption. PMID:11605533
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.
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)
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)
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
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)
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)
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)
PENELOPE DELTA, RECENTLY DISCOVERED WRITER
MALAPANI A.
2015-01-01
The aim of this article is to present a Greek writer, Penelope Delta. This writer has recently come up in the field of the studies of the Greek literature and, although thereare neither many translations of her works in foreign languages nor many theses or dissertations, she was chosen for the great interest for her works. Her books have been read by many generations, so she is considered a classical writer of Modern Greek Literature. The way she uses the Greek language, the unique characters...
FLUKA and PENELOPE simulations of 10keV to 10MeV photons in LYSO and soft tissue
Chin, M P W; Fassò, A; Ferrari, A; Ortega, P G; Sala, P R
2014-01-01
Monte Carlo simulations of electromagnetic particle interactions and transport by FLUKA and PENELOPE were compared. 10 key to 10 MeV incident photon beams impinged a LYSO crystal and a soft-tissue phantom. Central-axis as well as off-axis depth doses agreed within 1 s.d.; no systematic under- or overestimate of the pulse height spectra was observed from 100 keV to 10 MeV for both materials, agreement was within 5\\%. Simulation of photon and electron transport and interactions at this level of precision and reliability is of significant impact, for instance, on treatment monitoring of hadrontherapy where a code like FLUKA is needed to simulate the full suite of particles and interactions (not just electromagnetic). At the interaction-by-interaction level, apart from known differences in condensed history techniques, two-quanta positron annihilation at rest was found to differ between the two codes. PENELOPE produced a 511 key sharp line, whereas FLUKA produced visible acolinearity, a feature recently implemen...
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.
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)
Modeling the radio-induced effects in biological medium still requires accurate physics models to describe the interactions induced by all the charged particles present in the irradiated medium in detail. These interactions include inelastic as well as elastic processes. To check the accuracy of the very low energy models recently implemented into the GEANT4 toolkit for modeling the electron slowing-down in liquid water, the simulation of electron dose point kernels remains the preferential test. In this context, we here report normalized radial dose profiles, for mono-energetic point sources, computed in liquid water by using the very low energy “GEANT4-DNA” physics processes available in the GEANT4 toolkit. In the present study, we report an extensive intra-comparison of profiles obtained by a large selection of existing and well-documented Monte-Carlo codes, namely, EGSnrc, PENELOPE, CPA100, FLUKA and MCNPX. - Highlights: ► Normalized radial dose profiles are reported for mono-energetic electron sources. ► The low-energy “GEANT4-DNA” physics package is used for the calculations. ► A comparison with a large number of electron track-structure codes is proposed. ► Evident discrepancies in terms of shape and magnitude are reported. ► Accurate dose profiles have been provided by the GEANT4-DNA code
KAMCCO, a reactor physics Monte Carlo neutron transport code
KAMCCO is a 3-dimensional reactor Monte Carlo code for fast neutron physics problems. Two options are available for the solution of 1) the inhomogeneous time-dependent neutron transport equation (census time scheme), and 2) the homogeneous static neutron transport equation (generation cycle scheme). The user defines the desired output, e.g. estimates of reaction rates or neutron flux integrated over specified volumes in phase space and time intervals. Such primary quantities can be arbitrarily combined, also ratios of these quantities can be estimated with their errors. The Monte Carlo techniques are mostly analogue (exceptions: Importance sampling for collision processes, ELP/MELP, Russian roulette and splitting). Estimates are obtained from the collision and track length estimators. Elastic scattering takes into account first order anisotropy in the center of mass system. Inelastic scattering is processed via the evaporation model or via the excitation of discrete levels. For the calculation of cross sections, the energy is treated as a continuous variable. They are computed by a) linear interpolation, b) from optionally Doppler broadened single level Breit-Wigner resonances or c) from probability tables (in the region of statistically distributed resonances). (orig.)
Verification of Monte Carlo transport codes FLUKA, Mars and Shield
The present study is a continuation of the project 'Verification of Monte Carlo Transport Codes' which is running at GSI as a part of activation studies of FAIR relevant materials. It includes two parts: verification of stopping modules of FLUKA, MARS and SHIELD-A (with ATIMA stopping module) and verification of their isotope production modules. The first part is based on the measurements of energy deposition function of uranium ions in copper and stainless steel. The irradiation was done at 500 MeV/u and 950 MeV/u, the experiment was held at GSI from September 2004 until May 2005. The second part is based on gamma-activation studies of an aluminium target irradiated with an argon beam of 500 MeV/u in August 2009. Experimental depth profiling of the residual activity of the target is compared with the simulations. (authors)
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
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
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...
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
Configuration of the electron transport algorithm of PENELOPE to simulate ion chambers
Sempau, J.; Andreo, P.
2006-07-01
The stability of the electron transport algorithm implemented in the Monte Carlo code PENELOPE with respect to variations of its step length is analysed in the context of the simulation of ion chambers used in photon and electron dosimetry. More precisely, the degree of violation of the Fano theorem is quantified (to the 0.1% level) as a function of the simulation parameters that determine the step size. To meet the premises of the theorem, we define an infinite graphite phantom with a cavity delimited by two parallel planes (i.e., a slab) and filled with a 'gas' that has the same composition as graphite but a mass density a thousand-fold smaller. The cavity walls and the gas have identical cross sections, including the density effect associated with inelastic collisions. Electrons with initial kinetic energies equal to 0.01, 0.1, 1, 10 or 20 MeV are generated in the wall and in the gas with a uniform intensity per unit mass. Two configurations, motivated by the design of pancake- and thimble-type chambers, are considered, namely, with the initial direction of emission perpendicular or parallel to the gas-wall interface. This version of the Fano test avoids the need of photon regeneration and the calculation of photon energy absorption coefficients, two ingredients that are common to some alternative definitions of equivalent tests. In order to reduce the number of variables in the analysis, a global new simulation parameter, called the speedup parameter (a), is introduced. It is shown that setting a = 0.2, corresponding to values of the usual PENELOPE parameters of C1 = C2 = 0.02 and values of WCC and WCR that depend on the initial and absorption energies, is appropriate for maximum tolerances of the order of 0.2% with respect to an analogue, i.e., interaction-by-interaction, simulation of the same problem. The precise values of WCC and WCR do not seem to be critical to achieve this level of accuracy. The step-size dependence of the absorbed dose is explained in
Extension of PENELOPE to protons: Simulation of nuclear reactions and benchmark with Geant4
Purpose: Describing the implementation of nuclear reactions in the extension of the Monte Carlo code (MC) PENELOPE to protons (PENH) and benchmarking with Geant4.Methods: PENH is based on mixed-simulation mechanics for both elastic and inelastic electromagnetic collisions (EM). The adopted differential cross sections for EM elastic collisions are calculated using the eikonal approximation with the Dirac–Hartree–Fock–Slater atomic potential. Cross sections for EM inelastic collisions are computed within the relativistic Born approximation, using the Sternheimer–Liljequist model of the generalized oscillator strength. Nuclear elastic and inelastic collisions were simulated using explicitly the scattering analysis interactive dialin database for 1H and ICRU 63 data for 12C, 14N, 16O, 31P, and 40Ca. Secondary protons, alphas, and deuterons were all simulated as protons, with the energy adapted to ensure consistent range. Prompt gamma emission can also be simulated upon user request. Simulations were performed in a water phantom with nuclear interactions switched off or on and integral depth–dose distributions were compared. Binary-cascade and precompound models were used for Geant4. Initial energies of 100 and 250 MeV were considered. For cases with no nuclear interactions simulated, additional simulations in a water phantom with tight resolution (1 mm in all directions) were performed with FLUKA. Finally, integral depth–dose distributions for a 250 MeV energy were computed with Geant4 and PENH in a homogeneous phantom with, first, ICRU striated muscle and, second, ICRU compact bone.Results: For simulations with EM collisions only, integral depth–dose distributions were within 1%/1 mm for doses higher than 10% of the Bragg-peak dose. For central-axis depth–dose and lateral profiles in a phantom with tight resolution, there are significant deviations between Geant4 and PENH (up to 60%/1 cm for depth–dose distributions). The agreement is much better
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)
A special parallel plate ionization chamber, inserted in a slab phantom for the personal dose equivalent Hp(10) determination, was developed and characterized in this work. This ionization chamber has collecting electrodes and window made of graphite, and the walls and phantom made of PMMA. The tests comprise experimental evaluation following international standards and Monte Carlo simulations, employing the PENELOPE code to evaluate the design of this new dosimeter. The experimental tests were conducted employing the radioprotection level quality N-60 established at the IPEN, and all results were within the recommended standards. - Highlights: • A special ionization chamber, inserted in a slab phantom, was designed and evaluated. • This dosimeter was utilized for the Hp(10) determination. • The evaluation of this dosimeter followed international standards. • The PENELOPE Monte Carlo code was used to evaluate the design of this dosimeter. • The tests indicated that this dosimeter may be used as a reference dosimeter
Radiation therapy: dosimetry study of the effect of the composition of Pb alloys by PENELOPE
Jose McDonnell
2011-02-01
Full Text Available Radiotherapy is a widely used treatment for cancer. Currently applying the technique of Intensity Modulated Radiation Therapy, in which an important aspect is the modulation of the radiation beam to generate a non-uniform dose distribution in the tumor. One way to achieve the above non-uniform dose distribution is using solid compensators. In the market there are a number of materials used to manufacture compensators. Pb alloys on the market are: Cerromatrix, Rose, Wood, Newton, Darcet, whose compositions vary with respect to the composition of the lipowitz metal. This paper quantifies the dosimetric effects of the composition of commercial alloys, routinely used in radiotherapy. This quantification is important because of its impact on the total uncertainty of treatment accepted in the dosimetric calculations. To investigate the dosimetric effect of the composition of commercial alloys in the market we used the PENELOPE code, code that allows the simulation of radiation transport in different media by Monte Carlo method.The results show that there is a difference dosimetric respect lipowitz material, ranging from 7 % to 9 % for the materials investigated. These values indicate the importance of knowing exactly the dosimetric characteristics of the material used as compensator for their implications in the dose calculation.
Monte Carlo tools to supplement experimental microdosimetric spectra
Tissue-equivalent proportional counters (TEPCs) are widely used in experimental microdosimetry for characterising the radiation quality in radiation protection and radiation therapy environments. Generally, TEPCs are filled with tissue-equivalent gas mixtures, at low gas pressure, to simulate tissue site sizes similar to the cell nucleus (1 or 2 μm). The TEPC response using Monte Carlo (MC) codes can be applied to supplement experimental measurements. Most of general-purpose MC codes currently available recourse to the condensed-history approach to model the electron transport and do not transport low-energy electrons (60Co and 137Cs radiation at different simulated sizes (from 1.0 to 3.0 μm) in pure propane versus simulated spectra obtained with two general-purpose codes FLUKA and PENELOPE, which include a detailed simulation of electron-photon transport in arbitrary materials, including gases, is presented. A comparison between FLUKA and PENELOPE to generate 60Co and 137Cs microdosimetric spectra has been presented. Overall, calculated photon microdosimetric spectra and the experimental data showed a good correspondence. However, high differences are found for simulated sites of 1 μm. Maximum differences of ∼10 % are found between calculated and experimental data for yF-values, while there is an underestimation of yD-values of ∼4 and 8 % for FLUKA and PENELOPE, respectively, in comparison with experimental data. The agreement between calculated and experimental data is better for FLUKA than PENELOPE, despite the rougher approximations of the first code to model the electron transport. These calculations allow validating the range of applicability of multi-purpose MC codes, in microdosimetry applications. Within the range presented, simulated photon spectra could be employed to supplement microdosimetric spectra for simulated sites of >1 μm using FLUKA code. As suggested in a study regarding PENELOPE, to improve the TEPC-simulated response with PENELOPE
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:...
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
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.
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
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)
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)
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.
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.
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.
Aurora T: a Monte Carlo code for transportation of neutral atoms in a toroidal plasma
This paper contains a short description of Aurora code. This code have been developed at Princeton with Monte Carlo method for calculating neutral gas in cylindrical plasma. In this work subroutines such one can take in account toroidal geometry are developed
MCNP, a general Monte Carlo code for neutron and photon transport: 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
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
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)
Cell dosimetry is relevant regarding new efforts in specific molecular radiotherapy using Auger, CE and beta emitters. Absorbed dose in cells can be obtained by means of the dose per unit cumulated activity (S-values), together with the activity distribution. In this work, Monte Carlo simulation codes PENELOPE and MCNPX were used to obtain cellular S-values for point and extended sources of electrons and beta emitting radionuclides in the nucleus of breast (MDA-MB231, MCF7) and prostate (PC3) cancer cell models. - Highlights: • Cellular S-values were calculated using Penelope and MCPNX Monte Carlo codes. • S-values were obtained for e− and beta emitting radionuclides in cancer cell models. • Breast (MDA-MB231, MCF7) and prostate (PC3) cancer cell models were investigated. • Results are relevant for specific targeted molecular radiotherapy cell dosimetry
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
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)
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
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...
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)
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)
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
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)
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
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)
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
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
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.
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)
Extension of PENELOPE to protons: Simulation of nuclear reactions and benchmark with Geant4
Sterpin, E. [Center of Molecular Imaging, Radiotherapy and Oncology, Institut de recherche expérimentale et clinique, Université catholique de Louvain, Avenue Hippocrate 54, 1200 Brussels (Belgium); Sorriaux, J. [Center of Molecular Imaging, Radiotherapy and Oncology, Institut de recherche expérimentale et clinique, Université catholique de Louvain, Avenue Hippocrate 54, 1200 Brussels, Belgium and ICTEAM Institute, Université catholique de Louvain, Louvain-la-Neuve (Belgium); Vynckier, S. [Center of Molecular Imaging, Radiotherapy and Oncology, Institut de recherche expérimentale et clinique, Université catholique de Louvain, Avenue Hippocrate 54, 1200 Brussels, Belgium and Département de radiothérapie, Cliniques Universitaires Saint-Luc, Avenue Hippocrate 10, 1200 Brussels (Belgium)
2013-11-15
Purpose: Describing the implementation of nuclear reactions in the extension of the Monte Carlo code (MC) PENELOPE to protons (PENH) and benchmarking with Geant4.Methods: PENH is based on mixed-simulation mechanics for both elastic and inelastic electromagnetic collisions (EM). The adopted differential cross sections for EM elastic collisions are calculated using the eikonal approximation with the Dirac–Hartree–Fock–Slater atomic potential. Cross sections for EM inelastic collisions are computed within the relativistic Born approximation, using the Sternheimer–Liljequist model of the generalized oscillator strength. Nuclear elastic and inelastic collisions were simulated using explicitly the scattering analysis interactive dialin database for {sup 1}H and ICRU 63 data for {sup 12}C, {sup 14}N, {sup 16}O, {sup 31}P, and {sup 40}Ca. Secondary protons, alphas, and deuterons were all simulated as protons, with the energy adapted to ensure consistent range. Prompt gamma emission can also be simulated upon user request. Simulations were performed in a water phantom with nuclear interactions switched off or on and integral depth–dose distributions were compared. Binary-cascade and precompound models were used for Geant4. Initial energies of 100 and 250 MeV were considered. For cases with no nuclear interactions simulated, additional simulations in a water phantom with tight resolution (1 mm in all directions) were performed with FLUKA. Finally, integral depth–dose distributions for a 250 MeV energy were computed with Geant4 and PENH in a homogeneous phantom with, first, ICRU striated muscle and, second, ICRU compact bone.Results: For simulations with EM collisions only, integral depth–dose distributions were within 1%/1 mm for doses higher than 10% of the Bragg-peak dose. For central-axis depth–dose and lateral profiles in a phantom with tight resolution, there are significant deviations between Geant4 and PENH (up to 60%/1 cm for depth
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)
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.
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)
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)
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
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...
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)
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.
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
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)
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)
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.
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
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.
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)
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.
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
Platt, M. E.; Lewis, E. E.; Boehm, F.
1991-01-01
A Monte Carlo Fortran computer program was developed that uses two variance reduction techniques for computing system reliability applicable to solving very large highly reliable fault-tolerant systems. The program is consistent with the hybrid automated reliability predictor (HARP) code which employs behavioral decomposition and complex fault-error handling models. This new capability is called MC-HARP which efficiently solves reliability models with non-constant failures rates (Weibull). Common mode failure modeling is also a specialty.
The Serpent Monte Carlo Code: Status, Development and Applications in 2013
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
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.)
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
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
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)
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)
Chica, U. [Departamento de Física Atómica, Molecular y Nuclear, Universidad de Granada, E-18071 Granada, Spain and FISRAD S.A.S Carrera 64 a No 22-41, Bogotá D.C. (Colombia); Anguiano, M.; Lallena, A. M., E-mail: lallena@ugr.es [Departamento de Física Atómica, Molecular y Nuclear, Universidad de Granada, E-18071 Granada (Spain); Vilches, M. [Servicio de Radiofísica, Hospital Universitario “San Cecilio”, Avda. Dr. Olóriz, 16, E-18012 Granada (Spain)
2014-01-15
Purpose : To study the use of quality indexes based on ratios of absorbed doses in water at two different depths to characterize x-ray beams of low and medium energies. Methods : A total of 55 x-ray beam spectra were generated with the codes XCOMP5R and SPEKCALC and used as input of a series of Monte Carlo simulations performed with PENELOPE, in which the percentage depth doses in water and thek{sub Q,Q{sub 0}} factors, defined in the TRS-398 protocol, were determined for each beam. Some of these calculations were performed by simulating the ionization chamber PTW 30010. Results : The authors found that the relation betweenk{sub Q,Q{sub 0}} and the ratios of absorbed doses at two depths is almost linear. A set of ratios statistically compatible with that showing the best fit has been determined. Conclusions : The results of this study point out which of these ratios of absorbed doses in water could be used to better characterize x-ray beams of low and medium energies.
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
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
Monte Carlo simulations on a 9-node PC cluster
Monte Carlo simulation methods are frequently used in the fields of medical physics, dosimetry and metrology of ionising radiation. Nevertheless, the main drawback of this technique is to be computationally slow, because the statistical uncertainty of the result improves only as the square root of the computational time. We present a method, which allows to reduce by a factor 10 to 20 the used effective running time. In practice, the aim was to reduce the calculation time in the LNHB metrological applications from several weeks to a few days. This approach includes the use of a PC-cluster, under Linux operating system and PVM parallel library (version 3.4). The Monte Carlo codes EGS4, MCNP and PENELOPE have been implemented on this platform and for the two last ones adapted for running under the PVM environment. The maximum observed speedup is ranging from a factor 13 to 18 according to the codes and the problems to be simulated. (orig.)
Validation the Monte Carlo code RMC with C5G7 benchmark
Highlights: • The RMC code was verified based on the benchmark of C5G7. • Calculation speed of RMC is better than MCNP, especially in the flux tallies. • Eigenvalues calculated by RMC were within 2σ of the benchmark in all cases. • The pin by pin flux tallies of RMC are consistent with MCNP well. - Abstract: RMC (Reactor Monte Carlo code) is a new 3D Monte Carlo neutron transport code being developed by Department of Engineering Physics in Tsinghua University. The current version of RMC is a β version. In this paper, based on 2D and 3D benchmark of C5G7, the criticality calculation capacity of RMC was verified. Comparisons were made between the benchmark eigenvalues and those outputs by the RMC code. The RMC-generated eigenvalues were within two standard deviations of the benchmark and MCNP values in all cases. Additionally, the flux was compared pin by pin between MCNP and RMC. The flux tallies generated by RMC were found to be in well agreement with those from MCNP
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.
Optimization of Monte Carlo simulations
Bryskhe, Henrik
2009-01-01
This thesis considers several different techniques for optimizing Monte Carlo simulations. The Monte Carlo system used is Penelope but most of the techniques are applicable to other systems. The two mayor techniques are the usage of the graphics card to do geometry calculations, and raytracing. Using graphics card provides a very efficient way to do fast ray and triangle intersections. Raytracing provides an approximation of Monte Carlo simulation but is much faster to perform. A program was ...
Li, Junli; Li, Chunyan; Qiu, Rui; Yan, Congchong; Xie, Wenzhang; Wu, Zhen; Zeng, Zhi; Tung, Chuanjong
2015-09-01
The method of Monte Carlo simulation is a powerful tool to investigate the details of radiation biological damage at the molecular level. In this paper, a Monte Carlo code called NASIC (Nanodosimetry Monte Carlo Simulation Code) was developed. It includes physical module, pre-chemical module, chemical module, geometric module and DNA damage module. The physical module can simulate physical tracks of low-energy electrons in the liquid water event-by-event. More than one set of inelastic cross sections were calculated by applying the dielectric function method of Emfietzoglou's optical-data treatments, with different optical data sets and dispersion models. In the pre-chemical module, the ionised and excited water molecules undergo dissociation processes. In the chemical module, the produced radiolytic chemical species diffuse and react. In the geometric module, an atomic model of 46 chromatin fibres in a spherical nucleus of human lymphocyte was established. In the DNA damage module, the direct damages induced by the energy depositions of the electrons and the indirect damages induced by the radiolytic chemical species were calculated. The parameters should be adjusted to make the simulation results be agreed with the experimental results. In this paper, the influence study of the inelastic cross sections and vibrational excitation reaction on the parameters and the DNA strand break yields were studied. Further work of NASIC is underway. PMID:25883312
The method of Monte Carlo simulation is a powerful tool to investigate the details of radiation biological damage at the molecular level. In this paper, a Monte Carlo code called NASIC (Nanodosimetry Monte Carlo Simulation Code) was developed. It includes physical module, pre-chemical module, chemical module, geometric module and DNA damage module. The physical module can simulate physical tracks of low-energy electrons in the liquid water event-by-event. More than one set of inelastic cross sections were calculated by applying the dielectric function method of Emfietzoglou's optical-data treatments, with different optical data sets and dispersion models. In the pre-chemical module, the ionised and excited water molecules undergo dissociation processes. In the chemical module, the produced radiolytic chemical species diffuse and react. In the geometric module, an atomic model of 46 chromatin fibres in a spherical nucleus of human lymphocyte was established. In the DNA damage module, the direct damages induced by the energy depositions of the electrons and the indirect damages induced by the radiolytic chemical species were calculated. The parameters should be adjusted to make the simulation results be agreed with the experimental results. In this paper, the influence study of the inelastic cross sections and vibrational excitation reaction on the parameters and the DNA strand break yields were studied. Further work of NASIC is underway (authors)
Overview of TRIPOLI-4 version 7, Continuous-energy Monte Carlo Transport Code
The TRIPOLI-4 code is used essentially for four major classes of applications: shielding studies, criticality studies, core physics studies, and instrumentation studies. In this updated overview of the Monte Carlo transport code TRIPOLI-4, we list and describe its current main features, including recent developments or extended capacities like effective beta estimation, photo-nuclear reactions or extended mesh tallies. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. TRIPOLI-4 has support for execution in parallel mode. Special features and applications are also presented concerning: 'particles storage', resuming a stopped TRIPOLI-4 run, collision bands, Green's functions, source convergence in criticality mode, and mesh tally
Application of ENDF nuclear data for testing a Monte-Carlo neutron and photon transport code
A Monte-Carlo photon and neutron transport code was developed at OAEP. The code was written in C and C++ languages in an object-oriented programming style. Constructive solid geometry (CSG), rather than combinatorial, was used such that making its input file more readable and recognizable. As the first stage of code validation, data from some ENDF files, in the MCNP's specific format, were used and compared with experimental data. The neutron (from a 300 mCi Am/Be source) attenuation by water was chosen to compare the results. The agreement of the quantity 1/Σ among the calculation from SIPHON and MCNP, and the experiment - which are 10.39 cm, 9.71 cm and 10.25 cm respectively - was satisfactorily well within the experimental uncertainties. These results also agree with the 10.8 cm result of N.M., Mirza, et al. (author)
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.
Rabie, M.; Franck, C. M.
2016-06-01
We present a freely available MATLAB code for the simulation of electron transport in arbitrary gas mixtures in the presence of uniform electric fields. For steady-state electron transport, the program provides the transport coefficients, reaction rates and the electron energy distribution function. The program uses established Monte Carlo techniques and is compatible with the electron scattering cross section files from the open-access Plasma Data Exchange Project LXCat. The code is written in object-oriented design, allowing the tracing and visualization of the spatiotemporal evolution of electron swarms and the temporal development of the mean energy and the electron number due to attachment and/or ionization processes. We benchmark our code with well-known model gases as well as the real gases argon, N2, O2, CF4, SF6 and mixtures of N2 and O2.
Criticality studies in nuclear fuel cycle are based on Monte Carlo method. These codes use multigroup cross sections which can verify by experimental configurations or by use of reference codes such Tripoli 2. In this Tripoli 2 code nuclear data are errors attached and asked for experimental studies with critical experiences. This is one of the aim of this thesis. To calculate the keff of interacted fissile units we have used the multigroup Monte Carlo code Moret with convergence problems. A new estimator of reactions rates permit to better approximate the neutrons exchange between units and a new importance function has been tested. 2 annexes
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)
Analysing the statistics of group constants generated by Serpent 2 Monte Carlo code
An important topic in Monte Carlo neutron transport calculations is to verify that the statistics of the calculated estimates are correct. Undersampling, non-converged fission source distribution and inter-cycle correlations may result in inaccurate results. In this paper, we study the effect of the number of neutron histories on the distributions of homogenized group constants and assembly discontinuity factors generated using Serpent 2 Monte Carlo code. We apply two normality tests and a so-called “drift-in-mean” test to the batch-wise distributions of selected parameters generated for two assembly types taken from the MIT BEAVRS benchmark. The results imply that in the tested cases the batch-wise estimates of the studied group constants can be regarded as normally distributed. We also show that undersampling is an issue with the calculated assembly discontinuity factors when the number of neutron histories is small. (author)
This paper discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package developed and maintained at Oak Ridge National Laboratory. It has been developed to scale well from laptop to small computing clusters to advanced supercomputers. Special features of Shift include hybrid capabilities for variance reduction such as CADIS and FW-CADIS, and advanced parallel decomposition and tally methods optimized for scalability on supercomputing architectures. Shift has been validated and verified against various reactor physics benchmarks and compares well to other state-of-the-art Monte Carlo radiation transport codes such as MCNP5, CE KENO-VI, and OpenMC. Some specific benchmarks used for verification and validation include the CASL VERA criticality test suite and several Westinghouse AP1000® problems. These benchmark and scaling studies show promising results
The current basis for conversion coefficients for calibrating individual photon dosimeters in terms of dose equivalents is found in the series of papers by Grosswent. In his calculation the collision kerma inside the phantom is determined by calculation of the energy fluence at the point of interest and the use of the mass energy absorption coefficient. This approximates the local absorbed dose. Other Monte Carlo methods can be sued to provide calculations of the conversion coefficients. Rogers has calculated fluence-to-dose equivalent conversion factors with the Electron-Gamma Shower Version 3, EGS3, Monte Carlo program and produced results similar to Grosswent's calculations. This paper will report on calculations using the Integrated TIGER Series Version 3, ITS3, code to calculate the conversion coefficients in ICRU Tissue and in PMMA. A complete description of the input parameters to the program is given and comparison to previous results is included
Pandya, Tara M.; Johnson, Seth R.; Evans, Thomas M.; Davidson, Gregory G.; Hamilton, Steven P.; Godfrey, Andrew T.
2016-03-01
This work discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package authored at Oak Ridge National Laboratory. Shift has been developed to scale well from laptops to small computing clusters to advanced supercomputers and includes features such as support for multiple geometry and physics engines, hybrid capabilities for variance reduction methods such as the Consistent Adjoint-Driven Importance Sampling methodology, advanced parallel decompositions, and tally methods optimized for scalability on supercomputing architectures. The scaling studies presented in this paper demonstrate good weak and strong scaling behavior for the implemented algorithms. Shift has also been validated and verified against various reactor physics benchmarks, including the Consortium for Advanced Simulation of Light Water Reactors' Virtual Environment for Reactor Analysis criticality test suite and several Westinghouse AP1000® problems presented in this paper. These benchmark results compare well to those from other contemporary Monte Carlo codes such as MCNP5 and KENO.
An object-oriented implementation of a parallel Monte Carlo code for radiation transport
Santos, Pedro Duarte; Lani, Andrea
2016-05-01
This paper describes the main features of a state-of-the-art Monte Carlo solver for radiation transport which has been implemented within COOLFluiD, a world-class open source object-oriented platform for scientific simulations. The Monte Carlo code makes use of efficient ray tracing algorithms (for 2D, axisymmetric and 3D arbitrary unstructured meshes) which are described in detail. The solver accuracy is first verified in testcases for which analytical solutions are available, then validated for a space re-entry flight experiment (i.e. FIRE II) for which comparisons against both experiments and reference numerical solutions are provided. Through the flexible design of the physical models, ray tracing and parallelization strategy (fully reusing the mesh decomposition inherited by the fluid simulator), the implementation was made efficient and reusable.
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.)
Srna-Monte Carlo codes for proton transport simulation in combined and voxelized geometries
Ilic, R D; Stankovic, S J
2002-01-01
This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D) dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtaine...
ITS Version 6 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes.
Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William
2008-04-01
ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of lineartime-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 90. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.
Application of Monte Carlo method to build spectra library is useful to reduce experiment workload in Prompt Gamma Neutron Activation Analysis (PGNAA). The new Monte Carlo Code MOCA was used to simulate the response spectra of BGO detector for gamma rays from 137Cs, 60Co and neutron induced gamma rays from S and Ti. The results were compared with general code MCNP, show that the agreement of MOCA between simulation and experiment is better than MCNP. This research indicates that building spectra library by Monte Carlo method is feasible. (authors)
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.
Calculation of Gamma-ray Responses for HPGe Detectors with TRIPOLI-4 Monte Carlo Code
Lee, Yi-Kang; Garg, Ruchi
2014-06-01
The gamma-ray response calculation of HPGe (High Purity Germanium) detector is one of the most important topics of the Monte Carlo transport codes for nuclear instrumentation applications. In this study the new options of TRIPOLI-4 Monte Carlo transport code for gamma-ray spectrometry were investigated. Recent improvements include the gamma-rays modeling of the electron-position annihilation, the low energy electron transport modeling, and the low energy characteristic X-ray production. The impact of these improvements on the detector efficiency of the gamma-ray spectrometry calculations was verified. Four models of HPGe detectors and sample sources were studied. The germanium crystal, the dead layer of the crystal, the central hole, the beryllium window, and the metal housing are the essential parts in detector modeling. A point source, a disc source, and a cylindrical extended source containing a liquid radioactive solution were used to study the TRIPOLI-4 calculations for the gamma-ray energy deposition and the gamma-ray self-shielding. The calculations of full-energy-peak and total detector efficiencies for different sample-detector geometries were performed. Using TRIPOLI-4 code, different gamma-ray energies were applied in order to establish the efficiency curves of the HPGe gamma-ray detectors.
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)
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.
HERMES: a Monte Carlo Code for the Propagation of Ultra-High Energy Nuclei
De Domenico, Manlio; Lyberis, Haris; Settimo, Mariangela
2013-01-01
Although the recent experimental efforts to improve the observation of Ultra-High Energy Cosmic Rays (UHECRs) above $10^{18}$ eV, the origin and the composition of such particles is still unknown. In this work, we present the novel Monte Carlo code (HERMES) simulating the propagation of UHE nuclei, in the energy range between $10^{16}$ and $10^{22}$ eV, accounting for propagation in the intervening extragalactic and Galactic magnetic fields and nuclear interactions with relic photons of the e...
Sampling-Based Nuclear Data Uncertainty Quantification for Continuous Energy Monte Carlo Codes
Zhu, Ting
2015-01-01
The goal of the present PhD research is to establish a methodology of nuclear data uncertainty quantification (NDUQ) for MCNPX, the continuous-energy Monte-Carlo (M-C) code. The high fidelity (continuous-energy treatment and flexible geometry modelling) of MCNPX makes it the choice of routine criticality safety calculations at PSI/LRS, but also raises challenges for NDUQ by conventional sensitivity/uncertainty (S/U) methods. The methodology developed during this PhD research is fundamentally ...