Monte Carlo Radiation Transport Code Research Articles

Dosimetry analysis of two new design <SUP align="right">103</SUP>Pd interstitial brachytherapy sources

With increased demand for low 103 Pd (palladium) seed sources, to treat prostate and eye cancers, new sources have been designed and introduced. This paper presents the two new palladium brachytherapy sources, IR03- 103 Pd and IR04- 103 Pd that have been developed at Agricultural, Medical & Industrial Research School . The dosimetric parameters, such as the dose rate constant Λ , the radial dose function g ( r ) and the anisotropy function F ( r , θ ), around the sources have been characterised using Version 5 Monte Carlo radiation transport code in accordance with the update AAPM Task Group No. 43 Report (TG-43U1). The results indicated the dose rate constant of 0.689 ± 0.02 and 0.667 ± 0.02 cGy h -1 U -1 for the IR03- 103 Pd and IR04- 103 Pd sources, respectively, which are in acceptable agreement with other commercial seeds. The calculated results were compared with published results for those of other source manufacturers. However, they show an acceptable dose distribution, using for clinical applications is pending experimental dosimetry.

Dosimetry revisited for the HDR brachytherapy source model mHDR‐v2

Recently, the manufacturer of the HDR 192Ir mHDR-v2 brachytherapy source reported small design changes (referred to herein as mHDR-v2r) that are within the manufacturing tolerances but may alter the existing dosimetric data for this source. This study aimed to (1) check whether these changes affect the existing dosimetric data published for this source; (2) obtain new dosimetric data in close proximity to the source, including the contributions from 192Ir electrons and considering the absence of electronic equilibrium; and (3) obtain scatter dose components for collapsed cone treatment planning system implementation. Three different Monte Carlo (MC) radiation transport codes were used: MCNP5, PENELOPE2008, and GEANT4. The source was centrally positioned in a 40 cm radius water phantom. Absorbed dose and collision kerma were obtained using 0.1 mm (0.5 mm) thick voxels to provide high-resolution dosimetry near (far from) the source. Dose-rate distributions obtained with the three MC codes were compared. Simulations of mHDR-v2 and mHDR-v2r designs performed with three radiation transport codes showed agreement typically within 0.2% for r > or = 0.25 cm. Dosimetric contributions from source electrons were significant for r < 0.25 cm. The dose-rate constant and radial dose function were similar to those from previous MC studies of the mHDR-v2 design. The 2D anisotropy function also coincided with that of the mHDR-v2 design for r > or = 0.25 cm. Detailed results of dose distributions and scatter components are presented for the modified source design. Comparison of these results to prior MC studies showed agreement typically within 0.5% for r > or = 0.25 cm. If dosimetric data for r < 0.25 cm are not needed, dosimetric results from the prior MC studies will be adequate.

Microdosimetric assessment of the radiation quality of a therapeutic proton beam: comparison between numerical simulation and experimental measurements

Using protons for the treatment of ocular melanoma (especially of posterior pole tumours), the radiation quality of the beam must be precisely assessed to preserve the vision and to minimise the damage to healthy tissue. The radiation quality of a therapeutic proton beam at the Centre Antoine Lacassagne in Nice (France) was measured using microdosimetric techniques, i.e. a miniaturised version of a tissue-equivalent proportional counter. Measurements were performed in a 1-µm site at different depths in a Lucite phantom. Experimental data showed a significant increase in the beam quality at the distal edge of the spread-out Bragg peak (SOBP). In this paper, the numerical simulation of the experimental setup is done with the FLUKA Monte Carlo radiation transport code. The calculated microdosimetric spectra are compared with the measured ones at different depths in tissue for a monoenergetic proton beam (E=62 MeV) and for a modulated SOBP. Numerically and experimentally predicted relative biological effectiveness values are in good agreement. The calculated frequency-averaged and dose-averaged lineal energy mean values are consistent with measured data.

Helions at energies of 10 MeV to 1 TeV: conversion coefficients for fluence-to-absorbed dose, equivalent dose, effective dose and gray equivalent, calculated using Monte Carlo radiation transport code MCNPX 2.7.C

Conversion coefficients were calculated for fluence-to-absorbed dose, fluence-to-equivalent dose, fluence-to-effective dose and fluence-to-gray equivalent, for isotropic exposure of an adult male and an adult female to helions ((3)He(2+)) in the energy range of 10 MeV to 1 TeV (0.01-1000 GeV). Calculations were performed using Monte Carlo transport code MCNPX 2.7.C and BodyBuilder™ 1.3 anthropomorphic phantoms modified to allow calculation of effective dose using tissues and tissue weighting factors from either the 1990 or 2007 recommendations of the International Commission on Radiological Protection (ICRP), and gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. At 15 of the 19 energies for which coefficients for effective dose were calculated, coefficients based on ICRP 2007 and 1990 recommendations differed by less than 2%. The greatest difference, 62%, occurred at 100 MeV.

Gamma spectrometry inspection of TRIGA MARK II fuel using caesium isotopes

Gamma spectrometry is one of the common methods to inspect the spent fuel from research reactors. This method has been applied to in-pool measurements of the Spent Fuel Elements (SPEs) of the TRIGA Mark II research reactor. Due to mixed nature of the reactor core and complicated irradiation history of the fuel elements (FEs), the gamma spectrometry of the FE establishes improvements in the calculation and measurement of the SPE. In order to inspect the TRIGA SPE from dry storage and cooled fuel from the reactor pool, the selected spend fuels are scanned and measured using the fuel-scanning machine. Gamma spectrometry is performed by HPGe detector for spend fuel inspection and determination of the 137Cs activity and 134Cs/ 137Cs ratio. In this work, the steps of the detector calibration and the use of the Monte Carlo radiation transport code (MCNP5) have been described. In addition, the fuel-scanning machine and the gamma spectrometer are modelled by MCNP5 to simulate the gamma transport from fuel to detector. It also simulate the gamma spectrometer calibration for the burn up determination of the spend fuel. The results from MCNP5 simulation are applied to spectroscopic measurements and compared with the theoretical predictions of the neutronics code ORIGEN2 in this research work.

Tritons at energies of 10 MeV to 1 TeV: conversion coefficients for fluence-to-absorbed dose, equivalent dose, effective dose and gray equivalent, calculated using Monte Carlo radiation transport code MCNPX 2.7.C

Conversion coefficients were calculated for fluence-to-absorbed dose, fluence-to-equivalent dose, fluence-to-effective dose and fluence-to-gray equivalent for isotropic exposure of an adult female and an adult male to tritons ((3)H(+)) in the energy range of 10 MeV to 1 TeV (0.01-1000 GeV). Coefficients were calculated using Monte Carlo transport code MCNPX 2.7.C and BodyBuilder™ 1.3 anthropomorphic phantoms. Phantoms were modified to allow calculation of effective dose to a Reference Person using tissues and tissue weighting factors from 1990 and 2007 recommendations of the International Commission on Radiological Protection (ICRP) and calculation of gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. At 15 of the 19 energies for which coefficients for effective dose were calculated, coefficients based on ICRP 2007 and 1990 recommendations differed by less than 3%. The greatest difference, 43%, occurred at 30 MeV.

Deuterons at energies of 10 MeV to 1 TeV: conversion coefficients for fluence-to-absorbed dose, equivalent dose, effective dose and gray equivalent, calculated using Monte Carlo radiation transport code MCNPX 2.7.C

Conversion coefficients were calculated for fluence-to-absorbed dose, fluence-to-equivalent dose, fluence-to-effective dose and fluence-to-gray equivalent for isotropic exposure of an adult female and an adult male to deuterons ((2)H(+)) in the energy range 10 MeV-1 TeV (0.01-1000 GeV). Coefficients were calculated using the Monte Carlo transport code MCNPX 2.7.C and BodyBuilder™ 1.3 anthropomorphic phantoms. Phantoms were modified to allow calculation of the effective dose to a Reference Person using tissues and tissue weighting factors from 1990 and 2007 recommendations of the International Commission on Radiological Protection (ICRP) and gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. Coefficients for the equivalent and effective dose incorporated a radiation weighting factor of 2. At 15 of 19 energies for which coefficients for the effective dose were calculated, coefficients based on ICRP 1990 and 2007 recommendations differed by <3%. The greatest difference, 47%, occurred at 30 MeV.

Monte Carlo calculated TG‐60 dosimetry parameters for the emitter brachytherapy source

The formalism recommended by Task Group 60 (TG-60) of the American Association of Physicists in Medicine (AAPM) is applicable for beta sources. Radioactive biocompatible and biodegradable 153Sm glass seed without encapsulation is a beta- emitter radionuclide with a short half-life and delivers a high dose rate to the tumor in the millimeter range. This study presents the results of Monte Carlo calculations of the dosimetric parameters for the 153Sm brachytherapy source. Version 5 of the (MCNP) Monte Carlo radiation transport code was used to calculate two-dimensional dose distributions around the source. The dosimetric parameters of AAPM TG-60 recommendations including the reference dose rate, the radial dose function, the anisotropy function, and the one-dimensional anisotropy function were obtained. The dose rate value at the reference point was estimated to be 9.21 +/- 0.6 cGy h(-1) microCi(-1). Due to the low energy beta emitted from 153Sm sources, the dose fall-off profile is sharper than the other beta emitter sources. The calculated dosimetric parameters in this study are compared to several beta and photon emitting seeds. The results show the advantage of the 153Sm source in comparison with the other sources because of the rapid dose fall-off of beta ray and high dose rate at the short distances of the seed. The results would be helpful in the development of the radioactive implants using 153Sm seeds for the brachytherapy treatment.

A new radiation shielding material: Amethyst ore

This paper describes a new radiation shielding material, amethyst ore. We have determined the elemental composition of amethyst using WDXRF spectroscopy technique. To see the shielding capability of amethyst for several photon energies, these results have been used in simulation process by FLUKA Monte Carlo radiation transport code. Linear attenuation coefficients have been calculated according to the simulation results. Then, these values have been compared to a fine shielding concrete material. The results show that amethyst shields more gamma beams than concrete. This investigation is the first study about the radiation shielding properties of amethyst ore.

Monte Carlo simulations of the production of neutrons at iThemba LABS

Abstract FLUKA Monte Carlo radiation transport code has been used to simulate neutron fluence spectrum at iThemba LABS neutron beam facility. Neutron beams with energy up to 200 MeV can be produced using different targets such as 7 Li, 9 Be and 12 C bombarded with monoenergetic protons from the Separated-Sector Cyclotron. Simulated results at 66 MeV were compared with measured data. Different neutron emission angles with respect to the beam axis as well as the neutron background at different positions have been investigated.

Monte Carlo calculation of dosimetry parameters for the IR08‐ brachytherapy source

PurposeFor the treatment of some cancerous tumors using brachytherapy methods and low‐energy photon sources, such as and , the American Association of Physicists in Medicine Task Group No. 43U1 report recommends that the dosimetric parameters of a new brachytherapy source must be determined in two experimental and Monte Carlo theoretical methods before using each new source clinically. This study presents the results of Monte Carlo calculations of the dosimetric parameters for IR08‐ brachytherapy source design. IR08‐ seed has been manufactured at the Agricultural, Medical and Industrial Research School.MethodsVersion 5 of the (MCNP) Monte Carlo radiation transport code was used to calculate the dosimetry parameters around the source. Three geometric models of the seed, based on different locations of beads inside the titanium capsule, were simulated. The seed contains five resin beads of 0.6 mm diameter having uniformly absorbed in the bead volume, which were contained within a cylindrical titanium capsule having 0.8 mm outside diameter and 4.8 mm length.ResultsThe Monte Carlo calculated dose rate constant of the IR08‐ seed was found to be . Also in this study, the geometry function , line and point‐source radial dose functions and , and the anisotropy function , have been calculated at distances from 0.25 to 7 cm. The results of these calculations have been compared with measured values for an actual IR08‐ seed.ConclusionsThere are no statistical significant dosimetric differences among the three seed orientations in this study (i.e., ideal, vertical, and diagonal). However, the observed differences between the calculated and measured values could be explained by the measurement uncertainty and the configuration of the resin beads within the capsule and capsule orientation.

Simulation of image detectors in radiology for determination of scatter‐to‐primary ratios using Monte Carlo radiation transport codeMCNP/MCNPX

The purpose of this study was to compare and validate three methods to simulate radiographic image detectors with the Monte Carlo software MCNP/MCNPX in a time efficient way. The first detector model was the standard semideterministic radiography tally, which has been used in previous image simulation studies. Next to the radiography tally two alternative stochastic detector models were developed: A perfect energy integrating detector and a detector based on the energy absorbed in the detector material. Validation of three image detector models was performed by comparing calculated scatter-to-primary ratios (SPRs) with the published and experimentally acquired SPR values. For mammographic applications, SPRs computed with the radiography tally were up to 44% larger than the published results, while the SPRs computed with the perfect energy integrating detectors and the blur-free absorbed energy detector model were, on the average, 0.3% (ranging from -3% to 3%) and 0.4% (ranging from -5% to 5%) lower, respectively. For general radiography applications, the radiography tally overestimated the measured SPR by as much as 46%. The SPRs calculated with the perfect energy integrating detectors were, on the average, 4.7% (ranging from -5.3% to -4%) lower than the measured SPRs, whereas for the blur-free absorbed energy detector model, the calculated SPRs were, on the average, 1.3% (ranging from -0.1% to 2.4%) larger than the measured SPRs. For mammographic applications, both the perfect energy integrating detector model and the blur-free energy absorbing detector model can be used to simulate image detectors, whereas for conventional x-ray imaging using higher energies, the blur-free energy absorbing detector model is the most appropriate image detector model. The radiography tally overestimates the scattered part and should therefore not be used to simulate radiographic image detectors.

The Enhancements and Testing for the MCNPX 2.6.0 Depletion Capability

Monte Carlo–linked depletion methods have gained recent interest due to the ability to model complex three-dimensional geometries using continuous-energy cross sections. The integration of CINDER90 into the MCNPX Monte Carlo radiation transport code provides a completely self-contained Monte Carlo-linked depletion capability in a single Monte Carlo code that is compatible with most nuclear criticality (KCODE) particle tracking features in MCNPX. MCNPX depletion tracks all necessary reaction rates and follows as many isotopes as cross-section data permit. The objective of this work is (a) describe the MCNPX depletion methodology dating from the original linking of MONTEBURNS and MCNP to the first public release of the integrated capability (MCNPX 2.6.B, June 2006) that has been reported previously, (b) further detail the many new depletion capability enhancements since then leading to the present Radiation Safety Information Computational Center (RSICC) release, MCNPX 2.6.0, (c) report calculation results for the H. B. Robinson benchmark, and (d) detail new features available in MCNPX 2.7.A.Each version of MCNPX depletion starting from MCNPX 2.6.A leading to the official RSICC release of MCNPX 2.6.0 and the new beta release MCNPX 2.7.A included significant upgrades that addressed key issues from earlier versions. This paper details these key issues and the approach utilized to address the issues as enhancements for MCNPX 2.6.0. The MCNPX 2.6.0 depletion capability enhancements include (a) allowing the modeling of as large a system as computer memory capacity permits; (b) tracking every fission product available in ENDF/B VII.0; (c) enabling depletion in repeated structures geometries such as repeated arrays of fuel pins; (d) including metastable isotopes in burnup; and (e) manually changing the concentrations of any isotope during any time step by specified atom fraction, weight fraction, atom density, or gram density. These enhancements allow better detail to model the true system physics as well as to improve the robustness of the capability.H. B. Robinson benchmark calculations were completed to assess the validity of nuclide predictability of MCNPX 2.6.0. The results show comparisons of key actinide and fission products as compared to experiment and the SCALE-4 SAS2H sequence 27-group cross-section library (27BURNUPLIB) results. MCNPX 2.6.0 depletion results are within 4% of the experimental results for most major actinides.Two major depletion enhancements are available in the MCNPX 2.7.A beta release: improved 63-group flux querying and parallelization of the burnup interface routines in multiprocessor mode. Fixing the energy group querying routine does correctly tally the energy flux for use with isotopes not containing transport cross sections; however, results show <1% change in nuclide prediction for the benchmark test case. MCNPX 2.7.A parallelizes the depletion interface routines and running of CINDER90 so that different burnable regions of a given depletion system can be preprocessed, burned, and postprocessed on separate slave processors. The parallelization involves minimal communication between processors and therefore leads to significant computational performance enhancement.The combination of new enhancements and testing of the MCNPX 2.6.0 depletion computational system make this capability a valuable Monte Carlo-linked depletion tool. Additional testing and feature enhancements are under development to further improve the usefulness of the computational tool.

Intercomparision of Monte Carlo Radiation Transport Codes MCNPX, GEANT4, and FLUKA for Simulating Proton Radiotherapy of the Eye

Monte Carlo simulations of an ocular treatment beamline consisting of a nozzle and a water phantom were carried out using MCNPX, GEANT4, and FLUKA to compare the dosimetric accuracy and the simulation efficiency of the codes. Simulated central axis percent depth-dose profiles and cross-field dose profiles were compared with experimentally measured data for the comparison. Simulation speed was evaluated by comparing the number of proton histories simulated per second using each code. The results indicate that all the Monte Carlo transport codes calculate sufficiently accurate proton dose distributions in the eye and that the FLUKA transport code has the highest simulation efficiency.

Fluence to absorbed dose, effective dose and gray equivalent conversion coefficients for iron nuclei from 10 MeV to 1 TeV, calculated using Monte Carlo radiation transport code MCNPX 2.7.A

Conversion coefficients have been calculated for fluence-to-absorbed dose, fluence-to-effective dose and fluence-to-gray equivalent for isotropic exposure of an adult male and an adult female to (56)Fe(26+) in the energy range of 10 MeV to 1 TeV (0.01-1000 GeV). The coefficients were calculated using Monte Carlo transport code MCNPX 2.7.A and BodyBuilder 1.3 anthropomorphic phantoms modified to allow calculation of effective dose using tissues and tissue weighting factors from either the 1990 or 2007 recommendations of the International Commission on Radiological Protection (ICRP) and gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. Calculations using ICRP 2007 recommendations result in fluence-to-effective dose conversion coefficients that are almost identical at most energies to those calculated using ICRP 1990 recommendations.

Comparison of different MC techniques to evaluate BNCT dose profiles in phantom exposed tovarious neutron fields

The absorbed dose in BNCT (boron neutron capture therapy) consists of several radiation components with different physical properties and biological effectiveness. In order to assess the clinical efficacy of the beams, determining the dose profiles in tissues, Monte Carlo (MC) simulations are used. This paper presents a comparison between dose profiles calculated in different phantoms using two techniques: MC radiation transport code, MCNP-4C2 and BNCT MC treatment planning program, SERA (simulation environment for radiotherapy application). In this study MCNP is used as a reference tool. A preliminary test of SERA is performed using six monodirectional and monoenergetic beams directed onto a simple water phantom. In order to deeply investigate the effect of the different cross-section libraries and of the dose calculation methodology, monoenergetic and monodirectional beams directed toward a standard Snyder phantom are simulated. Neutron attenuation curves and dose profiles are calculated with both codes and the results are compared.

Skeletal contribution in lung measurements for the in vivo assessment of 241Am

Physical phantoms used for the calibration of in vivo measurement equipments represent an approximate geometry of the body and usually contain a uniform distribution of radionuclides in the organ of interest. Numerical voxel phantoms allow an improvement of the calibration method by means of a more realistic definition of the phantom and the source. The distribution of the activity in the tissues evolves with time and it can be estimated by solving the specific biokinetic model of the radionuclide. Those results of predicted activity retained in the compartments of the model can be translated into the organs of a voxel phantom to be used as the source of radiation for a Monte Carlo radiation transport code. In this way, different patterns of contamination, according to different times after the intake, can be modelled and appropriate detection efficiencies can be obtained. In particular, it is studied the influence of the skeletal contribution in lung measurements in the in vivo assessment of 241Am.

Alpha particles at energies of 10 MeV to 1 TeV: conversion coefficients for fluence-to-absorbed dose, effective dose, and gray equivalent, calculated using Monte Carlo radiation transport code MCNPX 2.7.A

Conversion coefficients have been calculated for fluence to absorbed dose, fluence to effective dose and fluence to gray equivalent, for isotropic exposure to alpha particles in the energy range of 10 MeV to 1 TeV (0.01-1000 GeV). The coefficients were calculated using Monte Carlo transport code MCNPX 2.7.A and BodyBuilder 1.3 anthropomorphic phantoms modified to allow calculation of effective dose to a Reference Person using tissues and tissue weighting factors from 1990 and 2007 recommendations of the International Commission on Radiological Protection (ICRP) and gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. Coefficients for effective dose are within 30 % of those calculated using ICRP 1990 recommendations.

Personal dose equivalent conversion coefficients for neutron fluence over the energy range of 20-250 MeV

Monte Carlo simulations were performed to extend existing neutron personal dose equivalent fluence-to-dose conversion coefficients to an energy of 250 MeV. Presently, conversion coefficients, H(p,slab)(10,alpha)/Phi, are given by ICRP-74 and ICRU-57 for a range of angles of radiation incidence (alpha = 0, 15, 30, 45, 60 and 75 degrees ) in the energy range from thermal to 20 MeV. Standard practice has been to base operational dose quantity calculations <20 MeV on the kerma approximation, which assumes that charged particle secondaries are locally deposited, or at least that charged particle equilibrium exists within the tally cell volume. However, with increasing neutron energy the kerma approximation may no longer be valid for some energetic secondaries such as protons. The Los Alamos Monte Carlo radiation transport code MCNPX was used for all absorbed dose calculations. Transport models and collision-based energy deposition tallies were used for neutron energies >20 MeV. Both light and heavy ions (HIs) (carbon, nitrogen and oxygen recoil nuclei) were transported down to a lower energy limit (1 keV for light ions and 5 MeV for HIs). Track energy below the limit was assumed to be locally deposited. For neutron tracks <20 MeV, kerma factors were used to obtain absorbed dose. Results are presented for a discrete set of angles of incidence on an ICRU tissue slab phantom.

Monte Carlo Calculations of Neutron Dose Conversion Coefficients for Reference Korean Male

Neutron fluence-to-organ dose conversion coefficients (DCCs) were calculated for a Reference Korean-adjusted male voxel model, HDRK-Man, for some standard irradiation geometries of external neutron beams using the Monte Carlo radiation transport code MCNPX. The calculated values were compared with those for other models. The results show that HDRK-Man provides consistent results for the DCC values considering the differences in the voxel models.