• 한국콘텐츠학회논문지
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• 제15권6호
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• pp.283-289
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• 2015

현재 방사선치료는 치료효과를 높이기 위해 고에너지 광자선의 사용이 증가하고 있는 추세이다. 일반적으로 6~8 MeV 이상의 고에너지 광자선을 사용하는 경우에는, 광핵반응에 의한 광중성자가 발생됨으로써 방사선 방호의 측면에서 많은 문제를 야기 시킬 수 있다. 이에 본 연구는 MCNPX를 이용하여 방사선 치료실의 광중성자 선량분포를 분석하였다. 그 결과 10 MV와 12 MV 구간에서 급격한 흡수선량의 증가를 보였다. 이를 통해 10 MV를 시작으로 광중성자 플루언스의 급격한 증가가 흡수선량으로 연계됨을 알 수 있었다. 또한 산출된 흡수선량을 바탕으로 등가선량을 환산한 결과는 ICRP 103 권고안의 경우, 낮은 에너지 범위에서 인체의 흡수선량에 대한 2차 광자의 기여를 반영함으로써 ICRP 60 권고안에 비해 낮은 등가선량을 나타냈다.

• 한국방사선학회논문지
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• 제13권2호
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• pp.241-246
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• 2019

4차 산업혁명의 기술 중 3D 프린팅 기술의 소재에 따른 선량평가를 통해 볼루스 적용 가능성을 평가하였다. 선량의 평가는 몬테카를로 방식의 MCNPX프로그램을 이용하였으며, 3D 프린트 물성은 ABS, PC, PLA 세 가지로 하였다. 그리하여 볼루스 10 mm와 동일한 효과를 보이는 두께를 산정한 결과 6 MeV 전자선의 경우 ABS 10 mm, PC 9 mm, PLA 9 mm로 나타났다. 6 MV X-선의 경우 ABS 11 mm, PC 10 mm, PLA 9 mm로 나타났다. 본 실험을 통해 3D 프린터 소재로 제작하는 조직등가물질이 볼루스를 대체할 수 있음을 확인할 수 있었다.

• 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.

• 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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• 제52권5호
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• pp.1002-1007
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• 2020

The purpose of the current study was to evaluate a spectrum formulation set employed to modify the neutron spectrum of D-D fusion neutrons in a IS plasma focus device using GEANT4, MCNPX2.6, and FLUKA codes. The set consists of a moderator, reflector, collimator and filters of fast neutron and gamma radiation, which placed on the path of 2.45 MeV neutron energy. The treated neutrons eliminate cancerous tissue with minimal damage to other healthy tissue in a method called neutron therapy. The system optimized for a total neutron yield of 10⁹ (n/s). The numerical results indicate that the GEANT4 code for the cubic geometry in the Beam Shaping Assembly 3 (BSA3) is the best choice for the energy of epithermal neutrons.

• Journal of Radiation Protection and Research
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• 제38권4호
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• pp.172-178
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• 2013

본 연구에서는 미래 파이로 시설에서의 핵물질 계량 연구를 위하여 개발하고 있는 통합비파괴측정장치(Unified Non-Destructive Assay, UNDA)의 중성자 검출 효율을 MCNPX 코드를 이용하여 평가하였다. 검출 효율 평가는 두 개의 다른 설계안의 UNDA에 대하여 수행되었으며, $^{252}Cf$ 중성자 발생 선원 위치에 따른 검출 효율 평가와 감손우라늄의 용기 두께 및 위치에 따른 검출 효율 평가를 수행하였다. $^{252}Cf$ 중성자 선원의 위치에 따른 UNDA의 검출 효율 결과는 6.83%부터 13.35%까지 분포로 나타났으며, $^{252}Cf$ 선원이 장치 내부의 상단에 위치할수록 검출 효율은 증가 후 감소하는 경향을 나타냈고, 선원이 외각에 위치될수록 효율이 증가하는 경향을 보였다. 감손우라늄 용기의 두께 및 위치에 따른 검출 효율 평가에서는 용기 두께가 증가할수록 검출 효율은 낮아지는 경향을 보이며, 용기 위치가 장치 상부에 위치될수록 효율은 감소하고, 외각에 위치할수록 효율은 증가하였다. 검출 효율은 $^{252}Cf$ 선원의 경우보다 약간 높게 나타났다(10.31~13.61%). 또한, 장치 상단에 고밀도 폴리에틸렌 덮개가 있는 설계안이 덮개가 없는 설계안 보다 평균적으로 약 2% 정도 중성자 검출 효율이 높은 것으로 평가되었다.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.269-274
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• 2019

$^{64}Cu$ is a favorable radionuclide in nuclear medicine applications because of its unique characteristics such as three types of decay (electron capture, ${\beta}^-$ and ${\beta}^+$) and 12.7 h half-life. Production of $^{64}Cu$ by irradiation $^{nat}Cu$ and $^{nat}CuNPs$ in Tehran Research Reactor was investigated. The characteristics of copper nanoparticles were investigated with SEM, TEM and XRD analysis. The cross section of $^{63}Cu(n,{\gamma})^{64}Cu$ reaction was done with TALYS-1.8 code. The activity value of $^{64}Cu$ was calculated with theoretical approach and MCNPX-2.6 code. The results were compared with related experimental results which showed good adaptations between them.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.151-158
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• 2018

The neutron powder diffractometer (NPD) is used to study a variety of technologically important and scientifically driven materials such as superconductors, multiferroics, catalysts, alloys, ceramics, cements, colossal magnetoresistance perovskites, magnets, thermoelectrics, zeolites, pharmaceuticals, etc. Monte Carlo-based codes are powerful tools to evaluate the neutronic behavior of the NPD. In the present study, MCNPX 2.6.0 and Vitess 3.3a codes were applied to simulate NPD facilities, which could be equipped with different optic devices such as pyrolytic graphite or neutron chopper. So, the Monte Carlo-based codes were used to simulate the NPD facility of the 5 MW Tehran Research Reactor. The simulation results were compared to the experimental data. The theoretical results showed good conformity to experimental data, which indicates acceptable performance of the Vitess 3.3a code in the neutron optic section of calculations. Another extracted result of this work shows that application of neutron chopper instead of monochromator could be efficient to keep neutron flux intensity higher than $10^6n/s/cm^2$ at sample position.

• 한국방사선학회논문지
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• 제11권4호
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• pp.197-203
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• 2017

의료용 선형가속장치의 두부 구성요소 중 광자 발생의 원인이 되는 타깃에 대한 연구로써, 타깃의 재질에 따른 광자를 분석하여 타깃 재질 별 발생하는 광자특성에 대한 기초자료를 제시하고자 한다. 본 연구에서는 몬테카를로 방식을 바탕으로 한 MCNPX를 사용하여 타깃 재질에 따른 6, 15 MV의 광자 특성을 비교분석하였다. 타깃 재질 별 평균에너지는 6 MV에서 1.69 ~ 1.84 MeV, 15 MV에서는 3.38 ~ 3.56 MeV로 분석되었다. Flux는 6 MV에서 $1.64{\times}10^{-5}{\sim}1.80{\times}10^{-5}{\sharp}/cm^2/e$, 15 MV는 $1.76{\times}10^{-4}{\sim}1.85{\times}10^{-4}{\sharp}/cm^2/e$로 계산되었다. 결과를 분석하면, 타깃 재질이 고원자번호일수록 평균에너지와 Flux가 증가하는 것으로 평가다. 본 연구를 바탕으로 광자의 물리적 특성에 대한 기초적인 자료를 제시할 수 있었으며, 추후 타깃 선정 시 경제성, 효율성은 물론 물리적 측면을 고려할 수 있어 적절한 선택을 할 수 있을 것으로 판단된다.

• 대한방사선기술학회지:방사선기술과학
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• 제41권2호
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• pp.141-148
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• 2018

This study examined the properties of photons and the dose distribution in a human body via a simulation where the total body irradiation(TBI) is performed on a pediatric anthropomorphic phantom and a child size water phantom. Based on this, we tried to find the optimal photon beam energy and material for beam spoiler. In this study, MCNPX (Ver. 2.5.0), a simulation program based on the Monte Carlo method, was used for the photon beam analysis and TBI simulation. Several different beam spoiler materials (plexiglass, copper, lead, aluminium) were used, and three different electron beam energies were used in the simulated accelerator to produce photon beams (6, 10, and 15 MeV). Moreover, both a water phantom for calculating the depth-dependent dosage and a pediatric anthropomorphic phantom for calculating the organ dosage were used. The homogeneity of photon beam was examined in different depths for the water phantom, which shows the 20%-40% difference for each material. Next, the org an doses on pediatric anthropomorphic phantom were examined, and the results showed that the average dose for each part of the body was skin 17.7 Gy, sexual gland 15.2 Gy, digestion 13.8 Gy, liver 11.8 Gy, kidney 9.2 Gy, lungs 6.2 Gy, and brain 4.6 Gy. Moreover, as for the organ doses according to materials, the highest dose was observed in lead while the lowest was observed in plexiglass. Plexiglass in current use is considered the most suitable material, and a 6 or 10 MV photon energy plan tailored to the patient condition is considered more suitable than a higher energy plan.