MCNPX Simulation Research Articles

Comparison between EGSnrc and MCNPX for X-ray target in 6 MV photon beam

The Monte Carlo codes, EGSnrc and MCNPX, have been specifically developed to study the particles transport in linear accelerator used in radiotherapy. A simple X-ray target 6 MV photon beam geometry was chosen to highlight similarities and differences of particles transport algorithms implemented in these Monte Carlo codes. The purpose of this study was to compare the particles transport in X-ray target 6 MV photon beam using EGSnrc and MCNPX codes. This target has 0.2484 cm thickness consisting of 0.0899 cm Tungsten and 0.1575 Copper. The global cut-off energies used in the simulations for electron and photon were 500 KeV and 10 KeV, respectively. Monte Carlo simulations were performed for 6.4 MeV monoenergetic beam. The parameters used in this simulation were using the DEFAULT setting. The particles fluence, spectral distribution after target and simulation time were analysed in this simulation. The results showed that the spectrum from EGSnrc and MCNPX was not well-matched. There was under 5% difference, especially in the build-up region. Meanwhile, the EGSnrc was 2 to 3 times more efficient than MCNPX code to achieve certain statistical precision. This happened because the differences of transport algorithm and approximation in these MC codes. EGSnrc just follow photons and electrons during simulation, while MCNPX can simulate transport 34 types of particles, including neutrons. It also causes the simulation time difference between these two codes. MCNPX simulation took longer than the EGSnrc. However, in some cases, differences in results as well as simulation time between 2 codes can be ignored.

Radiation damage assessment of the sixth SINQ target irradiation program based on MCNPX simulation

The 6th SINQ Target Irradiation Program (STIP VI) was conducted in SINQ Target 9 during 2011 and 2012. A radiation damage assessment based on a simulation using the MCNPX code was performed to obtain the irradiation parameters of displacement damage, helium and hydrogen concentrations in each specimen rod placed in the target.The gamma mapping performed on the proton beam entrance window after irradiation provided the proton fluence distribution used for implementing the source definition in the simulation. The proton and neutron fluences along the target as well as the energy deposition were calculated. Corresponding to the average beam current of 1.42 mA, the maximum neutron flux obtained is 5.95 × 1014 n/(cm2⋅ s) and the maximum energy deposition in the Zircaloy-2 cladding tubes is 568 W/cm3. The irradiation parameters of selected materials in each specimen rod were calculated by multiplying the proton and neutron fluences with the corresponding cross section data. The maximum displacement damage was obtained for Zircaloy-2 in Row 2 with a value of 72.1 dpa (displacement per atom). The maximum helium concentration was obtained for tantalum in Row 1 with a value of 3630 appm and the maximum hydrogen concentration was obtained for tungsten in Row 1 with a value of 17590 appm. All these values are the highest for the corresponding materials ever reached in a spallation target irradiation environment. The helium concentrations are also the highest ever obtained in bulk specimens.

Dosimetry of In Vivo Experiment for Lung Cancer Based on Boron Neutron Capture Therapy on Radial Piercing Beam Port Kartini Nuclear Reactor by MCNPX Simulation Method

Cancer is one of the leading causes of death globally, with lung cancer being among the most prevalent. Boron Neutron Capture Therapy (BNCT) is a cancer therapy method that uses the interaction between thermal neutrons and boron-10 which produces a decaying boron-11 particle and emits alpha, lithium 7 and gamma particles. A study was carried out to model an in vivo experiment of rat organisms that have lung cancer. Dimensions of a rat’s body were used in Konijnenberg research. Modeling lung cancer type, non-small cell lung cancer, was used in Monte Carlo N Particle-X. Lung cancer was modeled with a spherical geometry consisting of 3 dimensions: PTV, GTV, and CTV. In this case, the neutron source was from the radial piercing beam port of Kartini Reactor, Yogyakarta. The variation of boron concentration was 20, 25, 30, 35, 40, and 40 µg/g cancer. The output of the MCNP calculation was neutron scattering dose, gamma-ray dose and neutron flux from the reactor. A neutron flux was used to calculate the alpha proton and gamma-ray dose from the interaction of tissue material and thermal neutrons. The total dose was calculated from a four-dose component in BNCT. The results showed that the dose rate will increase when the boron concentration is higher, whereas irradiating time will decrease.

An investigation on shielding properties of BaO, MoO3 and P2O5 based glasses using MCNPX code

Abstract In the present work, some radiation shielding quantities (mass attenuation coefficients, effective atomic number, effective electron density, half value layer and mean free path) for various BaO–MoO3–P2O5 ternary glass systems have been determined within the 0.015–15 MeV energy range, using WinXCom program. Additionally, the mass attenuation coefficients of all the investigated glasses have been calculated using MCNPX simulation code (version 2.6.0) and compared to those of WinXCom results. Among the studied glasses, BaMoP8 glass sample with MoO3 content of 70% mol is found to have superior gamma-ray shielding characteristics. Moreover, the glasses studied in this paper possess better radiation shielding properties by providing shorter half value layer (HVL) than RS-253 G18 commercial glass and some concrete samples namely ordinary, hematite-serpentine and ilmanite-limonite.

Proton Therapy of eye using MCNPX code

Introduction: Proton radiotherapy is the one of advanced teletherapy methods. The protons deposit their maximum energy in a position called Bragg peak. Therefore, for treatment of cancer, the tumor should be placed at the Bragg peak or SOBP. The scattered photons and neutrons is a challenge in proton radiotherapy. The aim of this study is calculation of absorbed dose from scattered photon and neutron in eye proton therapy. Materials and Methods: In this study, the simulations were done using MCNPX simulation code. For this, a spherical tumor with 0.6 cm diameter was considered in the eye phantom. The eye phantom has been irradiated with proton beam of 50MeV. Tally F6 was used for dosimetry of scattered neutron and photons. Results: The results show that the absorbed dose per proton due to scattered neutron and photons in eye tissue are 8.817e-14 Gy and 1.653e-13 Gy respectively. Conclusion: In proton therapy of eye, the tumor can be treated accurately and the minimum absorbed dose and damage reaches the surrounding tissues. The results show that less than 1% of the total dose of protons is absorbed in the adjacent tissues.

Determination of internal absorbed dose from radiation of111In-Rituximab in body

Introduction: There are many ways to treat cancer and pain relief. One of the methods is radiopharmaceutical therapy. Indium 111-rituximab contains the radioisotope indium-111 and rituximab. The Indium-111, after decay, emits gamma rays with energies of 245 KeV and 171 KeV, and beta-radiation of 26 KeV. In this study, the absorbed dose of 111In-Rituximab in human body was calculated using mice data and MCNPX simulation code for the organs of heart, lung, liver, spleen, kidney, stomach, intestine and pancreas. Materials and Methods: Male phantom was used for this simulation. The source and target organs were defined in code and absorbed dose obtained for per particle with error of 5 percent. The source organs were determined as volume source in SDEF card of MCNP code. Simulations were done in two steps for gamma and beta sources and the final absorbed dose of each organ were multiplied by the amount of cumulated activity of 1MBq. Results: The results showed that the absorbed dose due to injection of 111In-Rituximab for organs of spleen and stomach are 2.2252E-02, 5.5979E-03 respectively. so these organs are at risk in treating cancer with this radionuclide. The errors for were less than 5 percent. Conclusion: Monte Carlo method and MCNPX simulation code are important methods for calculation of the absorbed dose after the distribution of rradionuclides in the body. Also the results show that two organs spleen and stomach are at risk in Indium 111-rituximab therapy.

Calculation of water equivalent ratios for various materials at proton energies ranging 10–500 MeV using MCNP, FLUKA, and GEANT4 Monte Carlo codes

Dosimetry of proton beams is generally evaluated in liquid water, or alternatively in solid phantoms via water equivalent ratios (WER). WER is defined as the ratio of proton range in liquid water to that in a phantom of certain material. Presently, WER is not available in the literature neither for a wide range of energies nor for variety of relevant materials. Thus, the goal of this study is to provide such data through Monte Carlo simulations. WER is calculated for 10–500 MeV energies for compact bone, adipose tissue, polymethyl methacrylate (PMMA), PTFE (teflon), graphite (C), aluminum (Al), copper (Cu), titanium (Ti), and gold (Au) using MCNPX.2.70, GEANT4, and FLUKA Monte Carlo (MC) codes. The MCNPX code was considered as the reference to which other codes were compared. The mean values of WER obtained through the MCNPX simulations for Au, Cu, Ti, Al, PTFE, graphite, PMMA, bone, and adipose tissue were 8.83, 5.40, 3.18, 2.03, 1.87, 1.52, 1.13, 1.71, and 0.96, respectively, for 10–500 MeV energy range. The maximum deviations of WER values between MCNPX and GEANT4 results were about 6.85% for adipose tissue at energies <20 MeV, whereas they were about 7.74%, 7.74% between MCNPX and FLUKA, for adipose and Al, respectively. This inter-code uncertainties are mainly due to different physic models and stopping powers in each code. Comparing the results to that in the literature, the range of discrepancy was found to be 0–8% with greatest discrepancy for Au. Based on the materials evaluated, the PMMA remained the closest to water, for a non-tissue solid material, with an average WER of 1.13, for proton energy ranging 10–500 MeV.

Radiation shielding parameters of some antioxidants using Monte Carlo method.

In this paper, radiation shielding parameters such as mass attenuation coefficients and half value layer (HVL) of some antioxidants are investigated using MCNPX (version 2.4.0). The validation of the generated MCNPX simulation geometry for antioxidant structures is provided by comparing the results with standard WinXcom data for radiation mass attenuation coefficients of antioxidants. Very good agreement between WİNXCOM and MCNPX was obtained. The results from the validated geometry were used to calculate the shielding parameters of different antioxidants. The radiation attenuation properties of each antioxidant were compared with each other. The results showed that, on average, the highest and the lowest radiation mass attenuation coefficients were observed on hesperidin and delphinidin chloride, respectively. It can be concluded that Monte Carlo simulation is a strong tool and an alternate method where experimental investigations are not possible and a standard simulation setup can be used in further studies for different biological structures. It can also be concluded that the obtained results from this study are very useful for radiology and radiotherapy applications where antioxidants are frequently used.

Photonuclear production of yttrium-88 – A high energy gamma emitter for hydrocarbon extraction applications

The use of fracing has risen over the past decade and revolutionized energy production in the US. However, there is still an impetus for further optimization of the extraction of oil and natural gas from vast shale reservoirs. In this work, we discuss photonuclear production of yttrium-88 as a promising radiotracer for fracing operations. Single neutron knock-out from natural monoisotopic yttrium-89 is an inexpensive process resulting in high activity of 88Y with minimal impurities. MCNPX simulations were performed to estimate the 88Y yield. Irradiations of natural yttrium using a 32 MeV electron linac equipped with a tungsten bremsstrahlung converter were done to benchmark the simulations. Activities of 88Y, 87gY, and 87mY were measured and found to be in good agreement with the predictions.

Comparative investigations of gamma and neutron radiation shielding parameters for different borate and tellurite glass systems using WinXCom program and MCNPX code

In the present article, for different chemical compositions of B2O3‒Bi2O3, B2O3‒Sb2O3, B2O3‒WO3‒La2O3, B2O3‒MoO3‒ZnO, and TeO2‒MO (M = Mg, Ba, and Zn) glasses, by applying WinXCom program we calculated the mass attenuation coefficient (μ/ρ) values, and from these values, the effective atomic number (Zeff), electron density (Ne), mean free path (MFP), half-value layer (HVL), and exposure buildup factor (EBF) values using Geometric progression (G‒P) fitting method, including macroscopic effective removal cross-section (ΣR) values for fast neutrons are evaluated for their potential applications as γ-ray and neutron radiation shielding materials. Moreover, the μ/ρ values of all the studied different glass compositions were computed using MCNPX simulation code and compared with WinXCom results. BBi65 glass has the highest μ/ρ, and Zeff values in the B2O3‒Bi2O3 glasses and lower values of MFP, HVL, and EBF. The maximum values of μ/ρ and Zeff are recorded for BSb70 in the B2O3‒ Sb2O3 glasses. It is found that the Zeff for B2O3‒Bi2O3 glasses is higher than those for B2O3‒ Sb2O3 glasses, which reveal that the B2O3‒Bi2O3 glasses have better shielding properties than the B2O3‒Sb2O3 glasses. The μ/ρ values of B2O3‒WO3‒La2O3 glasses are higher than those B2O3‒MoO3‒ZnO glasses, which indicate that B2O3‒WO3‒La2O3 glasses show preferable radiation shielding effectiveness comparing with B2O3‒MoO3‒ZnO glasses. The variation of different shielding parameters for the selected glasses was discussed according to the three photon interactions with matter (Photoelectric effect, Compton scattering, and pair production). The calculated μ/ρ and Zeff for the selected glasses have been compared with different glasses. HVL values are compared with ordinary, hematite-serpentine, ilmenite, steel-scrap and ilmenite-limonite concretes. It is found that ΣR values for B2O3‒Bi2O3, B2O3‒Sb2O3, B2O3‒WO3‒La2O3, B2O3‒MoO3‒ZnO, and TeO2‒MO (M = Mg, Ba, and Zn) glasses lie within the range 0.1312–0.2823 cm−1, 0.0876–0.0957 cm−1, 0.1180–0.1085 cm−1, 0.1066–0.1002 cm−1, and 0.1040–0.1075 cm−1, respectively.

A method to calculate the gamma ray detection efficiency of a cylindrical NaI (Tl) crystal

Given a wide range application of NaI(Tl) detector in industrial and medical sectors, computation of the related detection efficiency in different distances of a radioactive source, especially for calibration purposes, is the subject of radiation detection studies. In this work, a 2in both in radius and height cylindrical NaI (Tl) scintillator was used, and by changing the radial, axial, and diagonal positions of an isotropic 137Cs point source relative to the detector, the solid angles and the interaction probabilities of gamma photons with the detector's sensitive area have been calculated. The calculations present the geometric and intrinsic efficiency as the functions of detector's dimensions and the position of the source. The calculation model is in good agreement with experiment, and MCNPX simulation.

Reconstructive extremity dosimetry with 3D scanned hand model in MCNPX

We present a simulation method that can create accurate virtual models of hand with arbitrarily curved surfaces and perform distortion-free MCNPX simulations. Generally, MCNPX simulations of objects with arbitrarily curved surfaces are performed through voxelization. In this study, a polygon model is tetrahedralized by TetGen for the construction of the MCNPX geometry to be distortion-free. Then, dose estimation was successfully performed after converting the virtual model into an MCNPX input. A voxelized model was constructed for comparison purposes. The dose estimation functions of the two models were found to be similar, showing a similar amount of computing time by using the mesh tally option with 2e7 histories.

Selective contribution of concrete shielding on the equivalent and effective doses in a prostate radiotherapy treatment

Within a radiotherapy room, in addition to the primary beam, there is also secondary radiation that leaks from the accelerator head and scattering caused by surrounding objects, the patient’s body and even the walls of the shielded room itself, designed to protect the external individuals, disregarding effects on the patient. The aim of this work is to study the radiation effect on the patient’s healthy tissues caused by scattering from a concrete shield expressed as equivalent and effective doses. MCNPX simulations of the linear accelerator Varian 2100/2300C/D were performed for a MAX phantom of a patient in a typical radiotherapy room. In addition, calculations were made with and without shielding. It was concluded that the concrete shielding affects in 5% the effective dose absorbed by the patient.

The Effects of Image Quality due to Scattering X-ray according to increasing Patient Thickness

In this study, scattering factors affecting the quality of medical images were quantitatively analyzed and investigated. MCNPX simulation was conducted by using ANSI phantom, made of tissue equivalent materials, to calculate the scattering ratio occurred by the increase of the object thickness. Then, the result of the simulation was compared with the result of actual radiation measurement. In addition, we evaluated the image quality by the RMS evaluation, RSD and NPS analysis using X-ray images acquired with increasing object thickness. Furthermore, the scattering ratio was analyzed by increasing the thickness of acrylic phantom on chest phantom. The result showed that the scattering ratio was increased to 57.2%, 62.4%, and 66.8% from 48.9%, respectively, when the acrylic phantom thickness was increased by 1 inch from 6.1 inches. The results of MCNPX simulation and the actual measured scattering dose showed similar results. Also, as a result of RMS measurement from acquired x-ray images, the standard deviation decreased as the object thickness increased. However, in the RSD analysis considering the average incident dose, the results were increased from 0.028 to 0.039, 0.051, 0.062 as the acrylic phantom thickness was increased from 6.1 inches to 7.1 inch, 8.1 inch, and 9.1 inch, respectively. It can be seen that the increase of the scattering effect due to the increase of the object thickness reduces the SNR. Also, the NPS results obtained by measuring scattered radiation incident on the detector resulted in the increase of the noise as the object thickness increased.

DOSE ANALYZE OF BORON NEUTRON CAPTURE THERAPY (BNCT) AT SKIN CANCER MELANOMA USING MCNPX WITH NEUTRON SOURCE FROM THERMAL COLUMN OF KARTINI REACTOR

&lt;span&gt;This research aims to determine the amount of radiation dose rate that can be accepted and the irradiation time that is required from Boron Neutron Capture Therapy (BNCT) cancer therapy to treat melanoma skin cancer. This research used the simulation program, MCNPX by defining the geometric dimensions of the tissue component, and describing the radiation source that were used. The outputs obtained from the MCNPX simulation were the neutron flux and the neutron scattering dose that came out from the collimator. The value of neutron flux was used to calculate the dose which comes from the interaction between the neutron and the material in the cancer tissue. Based on the results of the research, the dose rate to treat cancer tissue for boron is 10 μg/g of tumor, which translates to about 0.019241 Gy/second and requires 25.98 minutes of irradiation time, 15 μg/g of tumor translates to 0.021854 Gy/second and requires 2.,87 minutes, 20 μg/g of tumor translates to 0.022902 Gy/second and requires 21,83 minutes, 25 μg/g of tumor translates to 0.0271275 Gy/second and requires 18.43 minutes, 30 μg/g of tumor translates to 0.0297658 Gy/second and requires 16.79 minutes, and 35 μg/g of tumor translates to 0.0343472 Gy/second and requires 14.55 minutes . The irradiation time needed for cancer tissue is shorter when boron concentration greater at the cancerous tissue.&lt;/span&gt;

CHARACTERIZATION OF FLUORESCENT NUCLEAR TRACK DETECTORS AS CRITICALITY DOSIMETERS II.

Fluorescent nuclear track detectors (FNTDs) as criticality dosimeters for both neutrons and gamma are further characterized in terms of angular dependence and quick dose assessment. The power spectrum integral depth profiles obtained from stacks of fluorescent images acquired within FNTDs exposed to a broad spectrum neutron field at various angles are analyzed to determine a calibration curve for angular dependence. MCNPX simulations were shown to be in good agreement with experimental results. A prototype triage reader was also designed and tested for quick assessment of dose. An unfolding technique incorporating both energy dependence and angular dependence is discussed. The advantages and shortcomings of using FNTDs in the event of a criticality excursion accident are analyzed.

DEVELOPMENT, CHARACTERIZATION AND TESTING OF A SIMPLIFIED BONNER SPHERE SYSTEM FOR RAPID NEUTRON FIELD CHARACTERIZATION.

Personal neutron dosimetry requires workplace specific correction factors from detailed neutron field characterization. A Bonner sphere system (BSS) with a 3He proportional counter and a 3 inch an a 9 inch sphere was developed at SCK·CEN for rapid neutron field characterization. The BSS energy and angular responses were simulated with MCNPX and validated by irradiations with a thermal neutron beam at the Belgian Reactor 1 (BR1) and 252Cf and Am-Be sources at the SCK·CEN Laboratory for Nuclear Calibration (LNK). The BSS unfolding performance was tested using the FRUIT unfolding code for measurements of the scattered neutron field at the LNK. The unfolded spectra were in good agreement with MCNPX simulations.

Experimental demonstration of a compact epithermal neutron source based on a high power laser

Epithermal neutrons from pulsed-spallation sources have revolutionised neutron science allowing scientists to acquire new insight into the structure and properties of matter. Here, we demonstrate that laser driven fast (∼MeV) neutrons can be efficiently moderated to epithermal energies with intrinsically short burst durations. In a proof-of-principle experiment using a 100 TW laser, a significant epithermal neutron flux of the order of 105 n/sr/pulse in the energy range of 0.5–300 eV was measured, produced by a compact moderator deployed downstream of the laser-driven fast neutron source. The moderator used in the campaign was specifically designed, by the help of MCNPX simulations, for an efficient and directional moderation of the fast neutron spectrum produced by a laser driven source.

Calculation of the low-energy gamma ray detection efficiency for a GM tube

Experiment and Monte-Carlo calculations indicate that the efficiency of the Geiger-Müller (GM) tube for gamma ray detection depends on the radial and axial positions of the point source. Given the extensive use that is made of this type of detector in industry and dosimetry, an accurate geometric model of the detector efficiency as a function of the source position would be welcome. In this paper, by calculating the solid angle of photons incident on the cathode wall and considering the probability of interaction of photons with the detector components, we obtain a calculation model for the detection efficiency of GM tubes (for soft gamma rays). This model is in good agreement with measurements and MCNPX simulations. We used isotropic 241Am as source.

Generation of the neutron response function of an NE213 scintillator for fusion applications

In this work we present a method to evaluate the neutron response function of an NE213 liquid scintillator. This method is particularly useful when the proton light yield function of the detector has not been measured, since it is based on a proton light yield function taken from literature, MCNPX simulations, measurements of gamma-rays from a calibration source and measurements of neutrons from fusion experiments with ohmic plasmas. The inclusion of the latter improves the description of the proton light yield function in the energy range of interest (around 2.46 MeV).We apply this method to an NE213 detector installed at JET, inside the radiation shielding of the magnetic proton recoil (MPRu) spectrometer, and present the results from the calibration along with some examples of application of the response function to perform neutron emission spectroscopy (NES) of fusion plasmas.We also investigate how the choice of the proton light yield function affects the NES analysis, finding that the result does not change significantly. This points to the fact that the method for the evaluation of the neutron response function is robust and gives reliable results.