Monte Carlo MCNPX Research Articles

A Monte Carlo approach to lung dose calculation in small fields used in intensity modulated radiation therapy and stereotactic body radiation therapy.

In the current study, the effect of electronic disequilibrium on lung dose with small photon beams was verified. The central axis absorbed dose in lung phantom was calculated by Monte Carlo (MC) method. The 6 and 18 MV photon beams of Varian Clinac 2100EX were simulated using MCNPX MC Code (Los Alamos national lab, USA). The MC model was used to calculate the depth doses water and low density water resembling the soft-tissue and lung, respectively. Four small field sizes including 0.5 cm2×0.5 cm2, 1 cm2×1 cm2, 2 cm2×2 cm2, and 3 cm2×3 cm2 were used in this study. Percentage of dose reduction in lung region relative to homogenous phantom for 6 MV photon beam were 44.6%, 39%, 13%, and 7% for 0.5 cm2×0.5 cm2, 1 cm2×1 cm2, 2 cm2×2 cm2, and 3 cm2×3 cm2 fields, respectively. For 18 MV photon beam, the results were found to be 82%, 69%, 46%, and 25.8% for the same field sizes, respectively. The solid tumor dose inside lung was reduced considerably between 17% and 35% for 18 MV beam, while there was only 9% dose reduction for tumor dose for 0.5 and 1 cm field sizes. Our study showed that the dose reduction with small fields in the lung was very enormous. Thus, inaccurate prediction of absorbed dose inside lung and also lung soft-tissue interfaces with small photon beams may lead to critical consequences for treatment outcome.

A Monte Carlo study on electron and neutron contamination caused by the presence of hip prosthesis in photon mode of a 15 MV Siemens PRIMUS linac

Several investigators have pointed out that electron and neutron contamination from high‐energy photon beams are clinically important. The aim of this study is to assess electron and neutron contamination production by various prostheses in a high‐energy photon beam of a medical linac. A 15 MV Siemens PRIMUS linac was simulated by MCNPX Monte Carlo (MC) code and the results of percentage depth dose (PDD) and dose profile values were compared with the measured data. Electron and neutron contaminations were calculated on the beam's central axis for Co‐Cr‐Mo, stainless steel, Ti‐alloy, and Ti hip prostheses through MC simulations. Dose increase factor (DIF) was calculated as the ratio of electron (neutron) dose at a point for 10×10 cm2 field size in presence of prosthesis to that at the same point in absence of prosthesis. DIF was estimated at different depths in a water phantom. Our MC‐calculated PDD and dose profile data are in good agreement with the corresponding measured values. Maximum dose increase factor for electron contamination for Co‐Cr‐Mo, stainless steel, Ti‐alloy, and Ti prostheses were equal to 1.18, 1.16, 1.16, and 1.14, respectively. The corresponding values for neutron contamination were respectively equal to: 184.55, 137.33, 40.66, and 43.17. Titanium‐based prostheses are recommended for the orthopedic practice of hip junction replacement. When treatment planning for a patient with hip prosthesis is performed for a high‐energy photon beam, attempt should be made to ensure that the prosthesis is not exposed to primary photons.PACS numbers: 87.56.bd, 87.55.kh, 87.55.Gh

Energy loss and online directional track visualization of fast electrons with the pixel detector Timepix

The spectral-, position- and tracking-response of Timepix to energetic electrons has been tested and characterized with well-defined low-intensity parallel beams of monoenergetic electrons in the 7–21 MeV range. The per-pixel energy sensitivity of Timepix serves as a position-sensitive dE/dx detector to determine the energy loss over along the particle track. The sampling path pitch can be set to 55 μm (pixel-size) and arbitrary values between 300 μm (sensor thickness) and nearly 2 mm. Timepix can register and visualize not only the position and trajectories but also the direction of trajectories and the rate of directional scattering of single fast electrons across the semiconductor sensor. The technique serves to measure the spatial distribution of a parallel beam, the beam size spread, transversal beam flux homogeneity and lateral straggling at the pixel-size scale. The mean scattering path along the beam axis and the mean path for lateral beam straggling in silicon can be determined. Measured energy loss distributions were compared with model and previous data as well as with dedicated Monte Carlo MCNPX simulations performed for this work. Tracking distributions were also simulated.

Monte Carlo simulation of electron modes of a Siemens Primus linac (8, 12 and 14 MeV)

AbstractBackgroundElectron mode is used for treatment of superficial tumours in linac-based radiotherapy.PurposeThe aim of present study is simulation of 8, 12 and 14 MeV electrons from a Siemens Primus linac using MCNPX Monte Carlo (MC) code and verification of the results based on comparison of the results with the measured data.Materials and methodsElectron mode for 8, 12 and 14 MeV electron energies of a Siemens Primus linac was simulated using MCNPX MC code. Percent depth dose (PDD) data for 10 × 10, 15 × 15 and 25 × 25 cm2 applicators obtained from MC simulations were compared with the corresponding measured data.ResultsGamma index values were less than unity in most of points for all the above-mentioned energies and applicators. However, for 25 × 25 cm2 applicator in 8 MeV energy, 10 × 10 cm2 applicator and 15 × 15 cm2 applicator in 14 MeV energy, there were four data points with gamma indices higher than unity. However among these data points, there are a number of cases with relatively large value of gamma index, these cases are positioned on the bremsstrahlung tail of the PDD curve which is not normally used in treatment planning.ConclusionThere was good agreement between the results of MC simulations developed in this study and the measured values. The obtained simulation programmes can be used in dosimetry of electron mode of Siemens Primus linac in the cases in which it is not easily feasible to perform experimental in-phantom measurements.

A new barite–colemanite concrete with lower neutron production in radiation therapy bunkers

The purpose of the current was to design a new high density concrete with low photoneutron production rate for radiation therapy bunker constructions. We used MCNPX Monte Carlo (MC) code for photoneutron productions in a radiation therapy bunker. The validated MC model of Varian Clinac 2100 C/D, 18MeV photon beam and a radiation therapy bunkers made of barite was used for photoneutron calculations. Colemanite mineral with weight fraction of 5% and 10% was added to the barite concrete in bunker walls. Photoneutrons absorbed doses at beam isocenter and at the maze entrance door were calculated. The Barite–colemanite concretes with two different compositions were built and their linear attenuation coefficients were measured against Cobalt-60, 9 and 18MeV photon beams. MC results showed that adding 5% and 10% Colemanite to Barite concrete reduces photoneutron production rate 7% and 16% respectively. However, experimental measurement of photon attenuation showed its lower photon attenuation coefficient and reduced mechanical properties. The results indicated that adding colemanite reduces photoneutrons received by the maze entrance door, but its photon attenuation and mechanical properties were compromised accordingly. Further studies on proposed concrete composition to enhance its mechanical properties are recommend.

A new analytical formula for neutron capture gamma dose calculations in double-bend mazes in radiation therapy

BackgroundPhotoneutrons are produced in radiation therapy with high energy photons. Also, capture gamma rays are the byproduct of neutrons interactions with wall material of radiotherapy rooms. AimIn the current study an analytical formula was proposed for capture gamma dose calculations in double bend mazes in radiation therapy rooms. Materials and methodsA total of 40 different layouts with double-bend mazes and a 18MeV photon beam of Varian 2100 Clinac were simulated using MCNPX Monte Carlo (MC) code. Neutron capture gamma ray dose equivalent was calculated by the MC method along the maze and at the maze entrance door of all the simulated rooms. Then, all MC resulted data were fitted to an empirical formula for capture gamma dose calculations. Wu–McGinley analytical formula for capture gamma dose equivalent at the maze entrance door in single-bend mazes was also used for comparison purposes. ResultsFor capture gamma dose equivalents at the maze entrance door, the difference of 2–11% was seen between MC and the derived equation, while the difference of 36–87% was found between MC and the Wu–McGinley methods. ConclusionOur results showed that the derived formula results were consistent with the MC results for all of 40 different geometries. However, as a new formula, further evaluations are required to validate its use in practical situations. Finally, its application is recommend for capture gamma dose calculations in double-bend mazes to improve shielding calculations.

Sensitization of the analytical methods for photoneutron calculations to the wall concrete composition in radiation therapy

The effect of wall material on photoneutron production in radiation therapy rooms was studied using Monte Carlo (MC) simulations. An analytical formula was proposed to take into account the concrete composition in photoneutron dose calculations. Using the MCNPX MC code, the 18 MV photon beam of the Varian Clinac 2100 and a typical treatment room with concrete compositions according to report No. 144 of National Council of Radiation Protection (NCRP) were simulated. Number of room produced photoneutrons per Gray of X-ray at the isocenter was determined for different types of concrete and named as “Q W ”. This new factor was inserted in the used formula for photoneutron fluence calculations at the inner entrance of maze. The photoneutron fluence was calculated using new proposed formula at the inner entrance of maze for all studied concretes. The difference between conventional and proposed equations varied from 11% to 46% for studied concretes. It was found that room produced photoneutrons could be significant for high density concretes. Additionally, applying the new proposed formula can consider the effect of wall material composition on the photoneutron production in high energy radiation therapy rooms. Further studies to confirm the accuracy of newly developed method is recommended. ► A radiation therapy room with a linac was simulated in the current study. ► Photoneutron production from different commercial concrete was studied. ► The photoneutron production varied from 11% to 46% dependent on concrete density and elemental composition. ► Photoneutron production was higher for high density concretes.

Monte Carlo calculations of positron emitter yields in proton radiotherapy

Positron emission tomography (PET) is a promising tool for monitoring the three-dimensional dose distribution in charged particle radiotherapy. PET imaging during or shortly after proton treatment is based on the detection of annihilation photons following the ß+-decay of radionuclides resulting from nuclear reactions in the irradiated tissue. Therapy monitoring is achieved by comparing the measured spatial distribution of irradiation-induced ß+-activity with the predicted distribution based on the treatment plan. The accuracy of the calculated distribution depends on the correctness of the computational models, implemented in the employed Monte Carlo (MC) codes that describe the interactions of the charged particle beam with matter and the production of radionuclides and secondary particles. However, no well-established theoretical models exist for predicting the nuclear interactions and so phenomenological models are typically used based on parameters derived from experimental data. Unfortunately, the experimental data presently available are insufficient to validate such phenomenological hadronic interaction models. Hence, a comparison among the models used by the different MC packages is desirable. In this work, starting from a common geometry, we compare the performances of MCNPX, GATE and PHITS MC codes in predicting the amount and spatial distribution of proton-induced activity, at therapeutic energies, to the already experimentally validated PET modelling based on the FLUKA MC code. In particular, we show how the amount of ß+-emitters produced in tissue-like media depends on the physics model and cross-sectional data used to describe the proton nuclear interactions, thus calling for future experimental campaigns aiming at supporting improvements of MC modelling for clinical application of PET monitoring.

Calculated angular responses of an RPL dosemeter to photon and beta radiation

The calculated dose equivalent response as a function of the angle has been examined for the radiophotoluminescent (RPL) glass dosemeter that was exposed to narrow series X-ray, N-60, N-80, N-100, N-150, N-200, N-250, N-300, photon sources ((60)Co and (137)Cs) and beta-ray emitter ((90)Sr/(90)Y) while mounted on an ISO water slab phantom. The angular dose equivalent responses H(p)(10) and H(p)(0.07) were calculated using the Monte Carlo MCNPX code. The RPL dosemeter and the phantom were rotated in the horizontal and vertical planes from a variety of angles of interest. The results were compared with the experimental data. Good agreement was found between the measured and calculated values of the relative dose equivalent angular responses of the RPL dosemeter.

First investigations of a silicon neutron detector with a carborane converter

We present a new design of a slow neutron detector, based on silicon detectors with innovative materials acting as conversion layers, plus preliminary results that will be the basis for a more complete study in the near future. The sensitive element of this detector is a Schottky barrier silicon diode covered with o-carborane, a novel boron-based converter material that detects neutrons by means of the 10B(n,α)7Li reaction. This sensor has been simulated with the MCNPX Monte-Carlo software in order to find the optimal converter layer thickness that maximizes the neutron detection efficiency. The simulated maximum efficiency is 2.7% for a 20 μm converter layer of the o-carborane fabricated with pure 10B and of 0.5% for the same compound but made with natural boron material. Moreover, the performance of test devices has been investigated by means of an 241AmBe neutron source and the results have been analyzed with the help of GEANT4 simulations.

Efficient Calculation of in vivo Efficiency Curves Using Variance Reduction Techniques

To optimize the in vivo lung monitoring of nuclear workers, realistic calibration coefficients are assessed using 3D anthropomorphic models and Monte Carlo simulations. In this study, a Livermore voxel phantom and the torso of the ICRP adult female reference voxel phantom were used. Monte Carlo MCNPX simulations were achieved to compute the calibration curve for a typical germanium counting system with a photon source of energy ranging from 15 keV to 1.4 MeV. However, photons of low energy are highly attenuated by body compounds while high energy photons are too penetrative to interact in the detectors. Hence, statistically insufficient counting efficiency values are obtained for both energy ranges. Thus, MCNPX variance reduction techniques were considered here to accelerate the simulations. The variance reduction techniques used were: the source biasing to favor low energy emissions; the emission direction biasing to select emission towards detectors; the forced collision to improve high energy photons interaction in germanium active cells. These techniques reduced the computing time by a factor of 20 and improved the associated statistical error with no significant variation in counting efficiency. Cell importance, mesh tally, weight windows and exponential transform variance reduction techniques were also tested to further accelerate computation.

Shielding Experiments at High Energy Accelerators of Fermilab (IV) —Calculation Analyses—

JASMIN - Japanese and American Study of Muon Interaction and Neutron detection - a program for studies of shielding and irradiation effect around high energy accelerators has been started until 2007 using high energy proton accelerators located in Fermi National Accelerator Laboratory (FNAL) as a collaboration of JAPAN and FNAL. The series of the presentations entitled Shielding experiments at high energy accelerators of Fermilab describes the part of the results of this collaboration regarding transport of secondary particles, neutron and muon, from 120 GeV proton induced reactions through experimental data and simulation. In this paper, calculation analyses using high energy particle transport Monte Carlo calculation codes, PHITS, MARS and MCNPX, for the experiments carried out at the anti-proton production target station were performed with a simplified two-dimensional geometry to validate the accuracy of spatial distribution of reaction rate in iron shield and the attenuation length. The calculation results were good agreement with the obtained experimental data for the spatial distribution. Neutron attenuation lengths of iron were also given by the data. Tendencies of the attenuation length of iron were found the PHITS and MARS values were larger and the MCNPX values were smaller than the data.

Simulating the angular response of Makrofol as a detector for neutron induced recoils

Abstract As a detector for neutron induced recoils, the response of Makrofol detectors is strongly dependent on the incident angle of neutrons. To investigate the effect of different incident angles, we performed simulations with different Monte Carlo codes, MCNPX and Geant4, and compared the results of the two codes. Simulated curves of the angular response were determined and experiments were carried out to verify the simulation results.

Evaluation of the dose and of the risk of cancer induction associated with the use of transmission X-ray body scanners using the Monte Carlo MCNPX code

In recent years, X–ray body scanners have been introduced at airports, penitentiaries and other places with considerable movement of people in order to combat drug trafficking, the entry of illegal materials and terrorism. However, although the application of this equipment in the national security area is indeed relevant, its use has caused a great deal of controversy, especially with regard to the doses absorbed and to the cancer induction risk associated with these exposures. The aim of this study is to use the Monte Carlo MCNPX code and the male adult voxel (MAX) and female adult voxel (FAX) phantoms, to evaluate the absorbed dose, effective dose and risk of cancer incidence attributable to exposures of individuals submitted to transmission X–ray body scanners in several projections. The effective dose values were calculated as recommended by the ICRP Publication 103 and the risk of cancer incidence were estimated through the BEIR VII Report.

Dosimetric impact of an brachytherapy source cable length modeled using a grid‐based Boltzmann transport equation solver

To evaluate the dose distributions of an 192Ir source (model VS2000) in homogeneous water geometry calculated using a deterministic grid-based Boltzmann transport equation solver (GBBS) in the commercial treatment planning system (TPS) (BRACHYVISION-ACUROS V8.8). Using percent dose differences (%deltaD), the GBBS (BV-ACUROS) was compared to the (1) published TG-43 data, (2) MCNPX Monte Carlo (MC) simulations of the 192Ir source centered in a 15 cm radius water sphere, and (3) TG-43 output from the TPS using vendor supplied (BV-TG43-vendor) and user extended (BV-TG43-extended) 2D anisotropy functions F(r, theta). BV-ACUROS assumes 1 mm of NiTi cable, while the TPS TG-43 algorithm uses data based on a 15 cm cable. MC models of various cable lengths were simulated. The MC simulations resulted in > 20% dose deviations along the cable for 1, 2, and 3 mm cable lengths relative to 15 cm. BV-ACUROS comparisons with BV-TG43-vendor and BV-TG43-extended yielded magnitude of differences, consistent with those seen in MC simulations. However, differences > 20% extended further (theta < or = 10 degrees) when using the vendor supplied anisotropy function F(ven0(r, theta). These differences were also seen in comparisons of F(r, theta) derived from the TPS output. The results suggest that %deltaD near the cable region is larger than previously estimated. The spatial distribution of the dose deviation is highly dependent on the reference TG-43 data used to compare to GBBS. The differences observed, while important to realize, should not have an impact on clinical dosimetry in homogeneous water.

Response studies of fission track detectors within moderators in CAVE A at GSI

Abstract At the Paul Scherrer Institut (PSI) fission track detectors are routinely used in moderators for determining the ambient dose equivalent in stray radiation fields due to neutrons. At the high-energy field CERF (CERN EU high-energy reference field) at CERN, the response of fission track detectors within two different moderators, where one moderator has an inlay of Pb to increase the response of the detector to high-energy neutrons, was surprisingly similar. To confirm this unexpected behaviour, a measurement campaign in another high-energy field, i.e. CAVE A at the Helmholtzzentrum fur Schwerionenforschung (GSI) was carried out. CAVE A at GSI is one of few facilities where a stray radiation field with neutrons above 20 MeV can be generated under controlled conditions. Reference doses were obtained by simulating the stray radiation field in the measurement positions. These simulations were carried out independently by GSI and PSI using two different Monte Carlo codes, i.e. Fluka and MCNPX, respectively. In this paper the measurements at GSI are described in detail. The results are then compared to previous measurement campaigns in the radiation field of CERF as well as 241 Am-Be and 252 Cf radionuclide sources, which all feature different neutron spectra. None of these campaigns confirm the improvement in sensitivity of the Pb enhanced moderator compared to the PSI standard moderator in combination with fission track detectors.

Photoneutron and capture gamma dose equivalent for different room and maze layouts in radiation therapy

In this paper the effect of treatment room and maze layout on the photoneutron and capture gamma dose equivalent in the maze was studied. MCNPX Monte Carlo (MC) code was used to simulate the Varian 2100 C/D Clinac 18 MV and four different room layouts. Two analytical methods, Wu-McGinley and McGinley, were used for dose calculations. The analytical methods overestimated the photoneutron dose (13-43 %) and gamma capture dose (16-95 %) comparing with the MC method at the maze entrance door. The results of MC method revealed that additional bend can cause a great reduction in photoneutron (5000 times) and capture gamma dose (50 times) in the maze entrance door.

Effect of wedge filter and field size on photoneutron dose equivalent for an 18 MV photon beam of a medical linear accelerator

Photoneutrons produced during radiation therapy with high energy photons is the main source of unwanted out-of-field received doses of patients. To analyze the neutron dose equivalent (NDE) for wedged beams and its variation with field size, Monte Carlo (MC) modeling of an 18 MV photon beam was performed using MCNPX MC code. The results revealed that the NDE is on average 6.5 times higher for wedged beams. For open beams, the NDE decreased with increasing field size especially for field sizes >20×20 cm 2. While, for wedged beams, the NDE increased with field size. It was suggested that the increase of NDE for wedged beams should be taken into account in radiation-induced secondary cancer risk estimations and radiation protection calculations.

Proton lateral broadening distribution comparisons between GRNTRN, MCNPX, and laboratory beam measurements

Recent developments in NASA’s deterministic High charge (Z) and Energy TRaNsport (HZETRN) code have included lateral broadening of primary ion beams due to small-angle multiple Coulomb scattering, and coupling of the ion-nuclear scattering interactions with energy loss and straggling. This new version of HZETRN is based on Green function methods, called GRNTRN, and is suitable for modeling transport with both space environment and laboratory boundary conditions. Multiple scattering processes are a necessary extension to GRNTRN in order to accurately model ion beam experiments, to simulate the physical and biological-effective radiation dose, and to develop new methods and strategies for light-ion radiation therapy. In this paper we compare GRNTRN simulations of proton lateral broadening distributions with beam measurements taken at Loma Linda University Proton Therapy Facility. The simulated and measured lateral broadening distributions are compared for a 250 MeV proton beam on aluminum, polyethylene, polystyrene, bone substitute, iron, and lead target materials. The GRNTRN results are also compared to simulations from the Monte Carlo MCNPX code for the same projectile-target combinations described above.

Comparison of Electron Dose-Point Kernels in Water Generated by the Monte Carlo Codes, PENELOPE, GEANT4, MCNPX, and ETRAN

Point kernels describe the energy deposited at a certain distance from an isotropic point source and are useful for nuclear medicine dosimetry. They can be used for absorbed-dose calculations for sources of various shapes and are also a useful tool when comparing different Monte Carlo (MC) codes. The aim of this study was to compare point kernels calculated by using the mixed MC code, PENELOPE (v. 2006), with point kernels calculated by using the condensed-history MC codes, ETRAN, GEANT4 (v. 8.2), and MCNPX (v. 2.5.0). Point kernels for electrons with initial energies of 10, 100, 500, and 1 MeV were simulated with PENELOPE. Spherical shells were placed around an isotropic point source at distances from 0 to 1.2 times the continuous-slowing-down-approximation range (R(CSDA)). Detailed (event-by-event) simulations were performed for electrons with initial energies of less than 1 MeV. For 1-MeV electrons, multiple scattering was included for energy losses less than 10 keV. Energy losses greater than 10 keV were simulated in a detailed way. The point kernels generated were used to calculate cellular S-values for monoenergetic electron sources. The point kernels obtained by using PENELOPE and ETRAN were also used to calculate cellular S-values for the high-energy beta-emitter, 90Y, the medium-energy beta-emitter, 177Lu, and the low-energy electron emitter, 103mRh. These S-values were also compared with the Medical Internal Radiation Dose (MIRD) cellular S-values. The greatest differences between the point kernels (mean difference calculated for distances, <0.9 r/R(CSDA)), using PENELOPE and those from ETRAN, GEANT4, and MCNPX, were 3.6%, 6.2%, and 14%, respectively. The greatest difference between the cellular S-values for monoenergetic electrons was 1.4%, 2.5%, and 6.9% for ETRAN, GEANT4, and MCNPX, respectively, compared to PENELOPE, if omitting the S-values when the activity was distributed on the cell surface for 10-keV electrons. The largest difference between the cellular S-values for the radionuclides, between PENELOPE and ETRAN, was seen for 177Lu (1.2%). There were large differences between the MIRD cellular S-values and those obtained from PENELOPE: up to 420% for monoenergetic electrons and <22% for the radionuclides, with the largest difference for 103mRh. In conclusion, differences were found between the point kernels generated by different MC codes, but these differences decreased when cellular S-values were calculated, and decreased even further when the energy spectra of the radionuclides were taken into consideration.