• Nuclear Engineering and Technology
• /
• 제52권12호
• /
• pp.2852-2859
• /
• 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 Radiation Protection and Research
• /
• 제42권3호
• /
• pp.154-157
• /
• 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
• /
• 제54권11호
• /
• pp.4231-4235
• /
• 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.

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

• Journal of Astronomy and Space Sciences
• /
• 제41권1호
• /
• pp.17-23
• /
• 2024

The safety of electronic components used in aerospace systems against cosmic rays is one of the most important requirements in their design and construction (especially satellites). In this work, by calculating the dose caused by proton beams in geostationary Earth orbit (GEO) orbit using the MCNPX Monte Carlo code and the MULLASSIS code, the effect of different structures in the protection of cosmic rays has been evaluated. A multi-layer radiation shield composed of aluminum, water and polyethylene was designed and its performance was compared with shielding made of aluminum alone. The results show that the absorbed dose by the simulated protective layers has increased by 35.3% and 44.1% for two-layer (aluminum, polyethylene) and three-layer (aluminum, water, polyethylene) protection respectively, and it is effective in the protection of electronic components. In addition to that, by replacing the multi-layer shield instead of the conventional aluminum shield, the mass reduction percentage will be 38.88 and 39.69, respectively, for the two-layer and three-layer shield compared to the aluminum shield.

• 전기전자학회논문지
• /
• 제19권3호
• /
• pp.282-287
• /
• 2015

본 연구에서는 연구에서는 Ce:GAGG 섬광체, 광섬유 그리고 광전자증배관을 이용하여 광섬유 검출기를 제작하였다. 섬광체의 단결정 크기는 MCNPX 코드를 이용하여 섬광체 깊이에 따른 감마선 계수효율을 전산모사하여 $3{\times}3{\times}20mm^3$로 설정하였다. 제작된 검출기는 표준 감마선원인 $^{137}Cs$과 $^{133}Ba$을 이용하여 세기의 따른 선형성 평가와 거리 변화에 따른 계수량 변화 측정을 하였다. 그 결과 추세선식 R-square 값이 0.99741로 매우 좋은 응답선형성을 보였고, 거리에 따른 검출 특성 또한 MCNPX값과 비교하였을 때 2% 이하로 좋은 검출특성을 보였다. 또한 단일선원과 혼합선원에서의 감마선 에너지 분광 결과 $^{137}Cs$은 662keV에서 그리고 $^{133}Ba$은 356keV에서 감마선 에너지 피크를 확인하였다. 광섬유 검출기를 사용한다면 작업자의 작업시간과 피폭을 줄여줄 것으로 보인다.

• 한국방사선학회논문지
• /
• 제8권5호
• /
• pp.249-254
• /
• 2014

의료용 선형가속장치는 1952년에 개발된 이후 방사선 치료에 사용되어 왔으며 그 활용도가 더욱 증가하고 있다. 현재는 6 MeV 이상의 광자 에너지를 사용하는 고 에너지 방사선치료가 보편화되어 사용되고 있으나, 광핵반응에 의한 중성자의 생성으로 환자 및 술자에 대한 피폭이 문제가 되고 있다. 이에 본 연구에서는 MCNPX를 사용하여 의료용 선형가속장치에서 발생되는 6~24 MV 광자선의 스펙트럼을 분석하고, 평균에너지 및 텅스텐의 중성자 생성 임계에너지인 7.41 MeV 이상의 광자 개수를 평가하였다. 그 결과 8 MV를 시작으로 24 MV에서는 전체 검출 광자 수에 비해 0.59%의 비율로서 광핵 반응을 일으킬 수 있는 광자수가 증가함을 알 수 있었다.

• Nuclear Engineering and Technology
• /
• 제54권12호
• /
• pp.4671-4678
• /
• 2022

The purpose of this comprehensive research is to observe the impact of scintillator crystal type on entire detection process. For this aim, MCNPX (version 2.6.0) is used for designing of a physical plastic scintillation detector available in our laboratory. The modelled detector structure is validated using previous studies in the literature. Next, different types of plastic scintillation crystals were assessed in the same geometry. Several fundamental detector properties are determined for six different plastic scintillation crystals. Additionally, the deposited energy quantities were computed using the MCNPX code. Although six scintillation crystals have comparable compositions, the findings clearly indicate that the crystal composed of PVT 80% + PPO 20% has superior counting and detecting characteristics when compared to the other crystals investigated. Moreover, it is observed that the highest deposited energy amount, which is a result of the highest collision number in the crystal volume, corresponds to a PVT 80% + PPO 20% crystal. Despite the fact that plastic detector crystals have similar chemical structures, this study found that performing advanced Monte Carlo simulations on the detection discrepancies within the structures can aid in the development of the most effective spectroscopy procedures by ensuring maximum efficiency prior to and during use.

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

• 한국안전학회지
• /
• 제33권2호
• /
• pp.28-31
• /
• 2018

In display and semi-conductor manufacturing process, there are numerous unstable factors such as particle concentration, minimal vibration, changes in magnetic field, or electrostatic that becomes an issue to be managed and controlled. In the recent, X-ray ionization is widely used that is neutralized by separating air or gas molecules in the area where the static must be resolved. The mono-type of X-ray ionizer was not capable to be used in $8^{th}$ generation panels manufacturing plant due to its insufficient ionizing coverage since the panel itself is approximately in $2m{\times}3m$. To resolve the current problem, the development of new type called, "Multi-type X-ray ionizer" has resulted in covering enough ionizing space in $8^{th}$ generation panels industry. Comparing mono and multi types with MCNPX code simulation, the multi one indicates more X-ray flux, efficiency, and ionization performance in comparison with either a mono-type or multi-type in array format. In addition, the ionizing efficiency of overlapping area with multi-type showed 30% higher effectiveness rate as to the ordinary mono-type.