• Journal of Radiation Protection and Research
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• v.42 no.3
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• pp.154-157
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• 2017

Background: Shielding properties of compound or mixture is presented in terms of mass attenuation coefficients using Monte Carlo simulation. Mass attenuation coefficients of cement, gypsum and the mixture of gypsum and $PbCO_3$ has been investigated using monte carlo MCNPX. Materials and Methods: The mass attenuation coefficients of cement, gypsum and the mixture of gypsum and $PbCO_3$ were calculated for photon energies 365.5, 661.6, 1,173.2, and 1,332.5 keV energies. Results and Discussion: The simulated values of mass attenuation coefficients were compared avaialable experimental results, theoretical values by XCOM and found good comparability of the results. Conclusion: Standard simulation geometry used in the present investigation would be very useful for various types of sample for shielding and dosimetry applications.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2852-2859
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• 2020

This paper deals with the simulation-based design optimization and experimental validation of the characteristics of an in-core silver Self-Powered Neutron Detector (SPND). Optimized dimensions of the SPND are determined by combining Monte Carlo simulations and analytical methods. As a first step, the Monte Carlo transport code MCNPX is used to follow the trajectory and fate of the neutrons emitted from an external source. This simulation is able to seamlessly integrate various phenomena, including neutron slowing-down and shielding effects. Then, the expected number of beta particles and their energy spectrum following a neutron capture reaction in the silver emitter are fetched from the TENDEL database using the JANIS software interface and integrated with the data from the first step to yield the origin and spectrum of the source electrons. Eventually, the MCNPX transport code is used for the Monte Carlo calculation of the ballistic current of beta particles in the various regions of the SPND. Then, the output current and the maximum insulator thickness to avoid breakdown are determined. The optimum design of the SPND is then manufactured and experimental tests are conducted. The calculated design parameters of this detector have been found in good agreement with the obtained experimental results.

• Journal of Korean Society for Quality Management
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• v.43 no.3
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• pp.239-252
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• 2015

Purpose: Lead has been widely used in radiation shielding for its low price and high workability. Recently in several europe countries, use of lead was banned for environmental issues. Also lead can cause health problems like alergies. Alternative materials for lead are highly required. The purpose of this study was to propose lead free radiation shielding material. Methods: Research of radiation shielding in Korea is not easy for certain limits such as radiation materials, experimental facilities and places. The collected data through the research were simulated using MCNPX. The simulation tools used for this study were utilized Monte Carlo method. Results: we suggest new design of lead free radiation shielding material using MCNPX code comparing shielding performance of new composite materials to lead. Conclusion: This newly introduced nano-scale composite of metal and polymer makes new chance for highly lightened radiation protective garments with endurable shielding performance.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3317-3323
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• 2022

Half Value Layer calculations theoretically need prior specification of linear attenuation calculations, since the HVL value is derived by dividing ln(2) by the linear attenuation coefficient. The purpose of this study was to establish a direct computational model for determining HVL, a vital parameter in nuclear radiation safety studies and shielding material design. Accordingly, a typical gamma-ray transmission setup has been modeled using MCNPX (version 2.4.0) general-purpose Monte Carlo code. The MCNPX code's INPUT file was designed with two detection locations for primary and secondary gamma-rays, as well as attenuator material between those detectors. Next, Half Value Layer values of some well-known gamma-ray shielding materials such as lead and ordinary concrete have been calculated throughout a broad gamma-ray energy range. The outcomes were then compared to data from the National Institute of Standards and Technology. The Half Value Layer values obtained from MCNPX were reported to be highly compatible with the HVL values obtained from the NIST standard database. Our results indicate that the developed INPUT file may be utilized for direct computations of Half Value Layer values for nuclear safety assessments as well as medical radiation applications. In conclusion, advanced simulation methods such as the Monte Carlo code are very powerful and useful instruments that should be considered for daily radiation safety measures. The modeled MCNPX input file will be provided to the scientific community upon reasonable request.

• The Journal of the Korea Contents Association
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• v.14 no.2
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• pp.467-474
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• 2014

The relative dose calculated by MCNPX and the relative dose measured by ionization chamber and solid phantoms evaluated the accuracy comparing with Monte Carlo simulation. In order to apply Monte Carlo simulation the intraluminal brachytherapy of extrahepatic bile duct cancer, 192Ir sealed radioactive source replicate, Bile duct and surrounding organs were made using KMIRD phantom based on a South Korea standard man. To check the absorbed dose of normal organs around bile duct, we set the specific effective energy and initial radioactivity to 1 Ci using MCNPX. Evaluation of the accuracy of the Monte Carlo simulation, the difference of the relative dose is the most 1.96% that satisfy the criteria that is the relative error less than 2% suggested by MCNPX code. In addition, The specific effective energy and absorbed dose of normal organs that were relatively adjacent to bile duct such as right side of kidney, liver, pancreas, transverse colon, spinal cord, stomach and small intestine were relatively high. on the contrary, the organs that were relatively distant to bile duct such as left side of kidney, spleen, ascending colon, descending colon and sigmoid colon were relatively low.

• Progress in Medical Physics
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• v.27 no.4
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• pp.232-235
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• 2016

The purpose of this study is to develop the treatment planning system (TPS) based on Monte-Carlo simulation for BNCT. In this paper, we will propose a method for dose estimation by Monte-Carlo simulation using the CT image, and will evaluate the accuracy of dose estimation of this TPS. The complicated geometry like a human body allows defining using the lattice function in MCNPX. The results of simulation such as flux or energy deposition averaged over a cell, can be obtained using the features of the tally provided by MCNPX. To assess the dose distribution and therapeutic effect, dose distribution was displayed on the CT image, and dose volume histogram (DVH) was employed in our developed system. The therapeutic effect can be efficiently evaluated by these evaluation tool. Our developed TPS could be effectively performed creating the voxel model from CT image, the estimation of each dose component, and evaluation of the BNCT plan.

• 대한방사선방어학회:학술대회논문집
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• 2011.11a
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• pp.188-189
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• 2011

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.785-794
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• 2016

The Argonne National Laboratory of the United States and the Kharkov Institute of Physics and Technology of the Ukraine have been collaborating on the design, development and construction of a neutron source facility at Kharkov Institute of Physics and Technology utilizing an electron-accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100-MeV electrons. The facility was designed to perform basic and applied nuclear research, produce medical isotopes, and train nuclear specialists. The biological shield of the accelerator building was designed to reduce the biological dose to less than 5.0e-03 mSv/h during operation. The main source of the biological dose for the accelerator building is the photons and neutrons generated from different interactions of leaked electrons from the electron gun and the accelerator sections with the surrounding components and materials. The Monte Carlo N-particle extended code (MCNPX) was used for the shielding calculations because of its capability to perform electron-, photon-, and neutron-coupled transport simulations. The photon dose was tallied using the MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is very small, ~0.01 neutron for 100-MeV electron and even smaller for lower-energy electrons. This causes difficulties for the Monte Carlo analyses and consumes tremendous computation resources for tallying the neutron dose outside the shield boundary with an acceptable accuracy. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were utilized for this study. The generated neutrons were banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron dose. The weight windows variance reduction technique was also utilized for both neutron and photon dose calculations. Two shielding materials, heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total dose outside the shield boundary less than 5.0e-03 mSv/h during operation. The shield configuration and parameters of the accelerator building were determined and are presented in this paper.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4231-4235
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• 2022

Considering the importance of deuterium in nuclear science including medical and industrial researches such as (BNCT) and nuclear reactors respectively, it is important to study various possible ways in addition to common methods for measuring its concentration. This study is an effort to measure deuterium concentration using PGNAA. The main idea is to calculate the area under 2.23 MeV gamma-rays photo peak resulting from neutron collision with Hydrogen atoms which are in mix with deuterium in samples. The study carried out by both simulation and experiment. Monte Carlo MCNPX2.6 code has been used for simulation and based on its acceptable results an experimental setup has been arranged. The coordination of results was in the range of R = 0.99 and R = 0.98 in simulation and experiment respectively. The accuracy of the study has been investigated by measuring the concentration of an unknown sample by both PGNAA and Fourier transform infrared spectroscopy (FT-IR) methods in which there were acceptable correlation between these two methods.

• Progress in Medical Physics
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• v.22 no.4
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• pp.190-197
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• 2011

Currently, the dose distribution calculation used by commercial treatment planning systems (TPSs) for high-dose rate (HDR) brachytherapy is derived from point and line source approximation method recommended by AAPM Task Group 43 (TG-43). However, the study of Monte Carlo (MC) simulation is required in order to assess the accuracy of dose calculation around three-dimensional Ir-192 source. In this study, geometry factor was calculated using segmented sources integration method by dividing microSelectron HDR Ir-192 source into smaller parts. The Monte Carlo code (MCNPX 2.5.0) was used to calculate the dose rate $\dot{D}(r,\theta)$ at a point ($r,\theta$) away from a HDR Ir-192 source in spherical water phantom with 30 cm diameter. Finally, anisotropy function and radial dose function were calculated from obtained results. The obtained geometry factor was compared with that calculated from line source approximation. Similarly, obtained anisotropy function and radial dose function were compared with those derived from MCPT results by Williamson. The geometry factor calculated from segmented sources integration method and line source approximation was within 0.2% for $r{\geq}0.5$ cm and 1.33% for r=0.1 cm, respectively. The relative-root mean square error (R-RMSE) of anisotropy function obtained by this study and Williamson was 2.33% for r=0.25 cm and within 1% for r>0.5 cm, respectively. The R-RMSE of radial dose function was 0.46% at radial distance from 0.1 to 14.0 cm. The geometry factor acquired from segmented sources integration method and line source approximation was in good agreement for $r{\geq}0.1$ cm. However, application of segmented sources integration method seems to be valid, since this method using three-dimensional Ir-192 source provides more realistic geometry factor. The anisotropy function and radial dose function estimated from MCNPX in this study and MCPT by Williamson are in good agreement within uncertainty of Monte Carlo codes except at radial distance of r=0.25 cm. It is expected that Monte Carlo code used in this study could be applied to other sources utilized for brachytherapy.