• Journal of Radiation Protection and Research
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• 제42권3호
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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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• 제52권12호
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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.

• 품질경영학회지
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• 제43권3호
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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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• 제54권9호
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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.

• 한국콘텐츠학회논문지
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• 제14권2호
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• pp.467-474
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• 2014

MCNPX를 통하여 계산한 상대선량과 고체팬텀과 전리함을 이용하여 측정한 상대선량을 비교하여 몬테카를로 시뮬레이션의 정확성을 평가하였다. 그리고 간외 담도암 관내근접방사선치료를 몬테카를로 시뮬레이션에 적용하기 위해 192Ir 밀봉방사성선원을 모사하였고, 한국 성인남성 표준인을 기초로 하는 KMIRD형 팬텀을 이용하여 담도 및 주변 장기를 제작하였다. 간외 담도암 관내근접방사선치료를 MCNPX를 이용하여 담도 주변 정상장기의 비유효에너지와 초기방사능을 1 Ci로 설정하여 흡수선량을 산정하였다. 몬테카를로 시뮬레이션의 정확성 평가에서 상대선량 차이가 가장 많은 지점이 1.96%로 MCNPX에서 제시한 상대오차 2%를 만족하는 것으로 나타났다. 또한, 담도 주변 정상장기의 비유효에너지 및 흡수선량은 담도와비교적 인접한 위치에 있는 우측신장, 간, 췌장, 횡행결장, 척수, 위장, 소장이 높았고, 담도와의 거리가 떨어져 있는 장기들인 좌측신장, 비장, 상행결장, 하행결장, S상결장이 낮게 나타났다.

• 한국의학물리학회지:의학물리
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• 제27권4호
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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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• 대한방사선방어학회 2011년도 추계 학술발표회 및 심포지엄
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• pp.188-189
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• 2011

• Nuclear Engineering and Technology
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• 제48권3호
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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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• 제54권11호
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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.

• 한국의학물리학회지:의학물리
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• 제22권4호
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• pp.190-197
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• 2011

고선량률 근접치료에 사용되는 상업용 선원과 치료계획 시스템들은 AAPM TG 43에서 권고하는 점 및 선 선원에 의해 선량분포를 계산한다. 하지만, 근접치료용 선원에 대한 인체 내의 정확한 선량계산을 위해서 3차원 부피의 선원을 고려하는 MC 기반의 선량계산 방법이 필요하다. 본 연구에서는 microSelectron HDR Ir-192 선원을 작은 부분으로 분할하여 계산하는 미소선원 적분법을 이용하여 기하학적 인수를 계산하였다. 또한, 범용 방사선 수송코드인 MCNPX를 사용하여 30 cm 직경의 구형 물 팬텀 내에서 선원의 선량률을 계산하여 비등방성함수와 반경선량함수를 구하였다. 그 결과를 MC 기반 광자 수송코드인 MCPT를 사용하여 계산한 Williamson의 결과와 비교 및 분석하였다. 미소선원 적분법과 선 선원 근사법에 따른 기하학적 인수는 $r{\geq}0.5cm$에서는 0.2% 이내에서 일치하였고 r=0.1 cm일 때 1.33%의 차이를 보였다. 본 연구에서 계산된 비등방성함수와 반경선량함수가 Williamson의 계산된 결과의 차이는 비등방성함수의 경우 r=0.25 cm에 서 2.33%의 가장 큰 R-RMSE를 보였고 $r{\geq}0.5cm$에서는 1% 미만의 R-RMSE를 보였다. 반경선량함수의 경우는 r=0.1~14.0 cm에서 0.46%의 R-RMSE를 보였다. 미소선원 적분법과 선 선원 근사법으로 계산한 기하학적 인수는 $r{\geq}0.1cm$에서 잘 일치하지만 3차원의 Ir-192 선원을 적용하여 계산한 미소선원 적분법이 실제 기하학적 인수를 잘 반영할 것으로 생각된다. r=0.25 cm에서 비등방성함수를 제외하고는 MCPT와 MCNPX의 몬테칼로 코드를 이용하여 얻어진 비등방성함수와 반경선량함수는 각각의 몬테칼로 코드에 대한 불확실성 이내에서 잘 일치함을 확인하였다. 따라서 MCNPX 전산모사 결과를 통해 TG-43의 선량 계산식에 사용된 인자를 Williamson 등의 결과와 비교 및 검증함으로써, 추후 다른 종류의 선원에 대해서도 Monte Carlo 기반의 연구가 가능할 것으로 기대된다.