• 한국산학기술학회논문지
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• 제22권5호
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• pp.530-536
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• 2021

본 논문에서는 전자기기 등에서 전자파 차폐 및 방열소재로 많이 채택되는 동 메쉬 시트와 천연그라파이트 시트를 감압 접착제없이 합지시켜 개발된 차폐 및 방열의 기능을 동시에 가지는 하이브리드시트의 성능 측정결과를 제시하였다. 객관적인 방열 성능을 확인하기 위해 2개의 다른 제품들과 수직 및 수평 열전도도를 각각 측정하여 결과를 비교하였으며, 전자파 차폐 성능은 CISPR 11규격에 따른 복사방출시험을 3m 전자파 무향실에서 진행하여 확인하였다. 수직 열전도도의 경우 제안된 하이브리드 시트가 방열코팅이 된 알루미늄 시트 대비 약 8.63배, 감압 접착제로 인조 그라파이트를 합지시킨 구리 시트에 비해 18.7배 높은 수준이였으며, 수평 열전도도는 인조 그라파이트를 합지시킨 구리 시트에 비해 약 0.64배, 방열 코팅된 알루미늄 시트에 대해서는 약 1.76배로 나타났다. 동일한 열원에서 각 시트들을 적용 후 측정한 결과에서는 제안된 하이브리드 시트가 열방출 기능이 가장 우수하였고 복사방출시험에서는 방사노이즈들이 상당 부분 제거되는 결과를 얻었다.

• Journal of Radiation Protection and Research
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• 제41권2호
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• pp.111-115
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• 2016

Background: The scattered photons cause reduction of the contrast of radiographic image and it results in the degradation of the quality of the image. In order to acquire better quality image, an anti-scattering x-ray gird should be equipped in radiography system. Materials and Methods: The X-ray anti-scattering grid of the inclined type based on the hybrid concept for that of parallel and focused type was tested by MCNP code. The MCNPX 2.7.0 was used for the simulation based test. The geometry for the test was based on the IEC 60627 which was an international standard for diagnostic X-ray imaging equipment-Characteristics of general purpose and mammographic anti-scatter grids. Results and Discussion: The performance of grids with four inclined shielding material types was compared with that of the parallel type. The grid with completely tapered type the best performance where there were little performance difference according to the degree of inclination. Conclusion: It was shown that the grid of inclined type had better performance than that of parallel one.

• 한국전자파학회논문지
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• 제14권9호
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• pp.950-954
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• 2003

본 논문에서는 내부 전자파원에 의해 여기된 공동의 내벽 한 면에 손실 유전체를 설치한 경우 공동 내부로 공급된 전력과 반사계수를 계산하였으며, 손실 유전체에 의해 공동 내부의 전자계가 저감되는 특성을 검토하고있다. 이론 해석으로는 내부 전자파원의 전류 분포 및 손실 유전체 경계면에서의 전계분포에 관한 연립 적분방정식을 유도하고 Galerkin의 모멘트법으로 해석하여 공동체 공급된 전력과 반사계수를 구하고 있다. 이론해석 결과, carbon을 함유한 발포 폴리스티렌을 손실 유전체로 사용하여 유전체의 두께와 carbon 함유량을 조절함으로써 공동 내부 전자파원으로부터의 전자파 방사를 저감시킬 수 있음을 보이고 있으며, 공급 전력의 실험치와도 비교하여 이론 해석의 타당성을 확인하고 있다.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4130-4136
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• 2021

By polymerization of gadolinium methacrylate (Gd (MAA)₃), lead methacrylate (Pb(MAA)₂) and methyl methacrylate (MMA), Gd and Pb were chemically bonded into polymers. The X-ray shielding performance was evaluated by Monte Carlo simulation method, and the results showed that the more metal functional organic monomer, the better the shielding performance of terpolymers. When the X-ray energy is 65 keV, Gd (MAA)₃-containing polymers have better shielding performance than Pb(MAA)₂-containing polymers. Gd could compensate for the weak absorption region of Pb. Therefore, polymers containing both Gd and Pb enhanced shielding efficiency against X-ray in various low-energy ranges. For obtaining terpolymers with uniform monomer compositions, the relationship between the monomer composition of the terpolymers and the conversion level was optimized by calculating the reactivity ratios. The value of reactivity ratios of r (Gd (MAA)₃/Pb(MAA)₂), r (Pb(MAA)₂/Gd (MAA)₃), r (Gd (MAA)₃/MMA), r (MMA/Gd (MAA)₃), r (Pb(MAA)₂/MMA) and r (MMA/Pb(MAA)₂) was 0.483, 0.004, 0.338, 2.508, 0.255, 0.029. The terpolymers with uniform monomer composition could be obtained by controlling the monomer compositions or conversion levels. The results can provide new radiation protection materials and contribute to the improvement in nuclear safety.

• 방사성폐기물학회지
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• 제5권2호
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• pp.109-112
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• 2007

나노 규모의 1차원 양자 구조물을 이용한 전자선 차폐 가능성에 관한 이론적 배경과 한계를 정리한다. 나노 구조물을 이용한 전자선 차폐는 차폐재의 경량화와 소형화에 크게 기여할 것으로 예상되나, 실용화를 위해서는 아직 연구되어야 할 분야가 많다. 임의의 1차원 포텐셜 장벽을 대상으로 양자투과계수 계산을 실행하여, 나노 구조물의 전자선 차폐와 관련된 문제점들을 살펴본다.

• 대한방사선기술학회지:방사선기술과학
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• 제38권3호
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• pp.237-244
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• 2015

핵의학 종사자는 PET/CT 업무 환경 중 방사성 의약품 취급 시 상당히 높은 피폭선량을 받는다고 알려져 있으며, 이를 최소화하기 위해 적절한 차폐기구의 사용이 요구된다. 이에 본 연구에서는 몬테카를로 기법을 기반으로 한 모의실험과 실측을 통해 18F-FDG 선원 취급 시 Apron 착용에 대한 차폐효과에 대해 분석하였다. 그 결과, 모의실험의 경우 선원의 취급 위치에 따라 인체 장기별 선량 분포가 각각 다른 양상을 나타냈고, Apron 납 두께별 선량 감소율은 선원과 장기와의 위치가 근접할수록, 선원과의 접촉 거리가 멀수록 낮은 경향을 나타냈다. 선량 측정 장비를 통한 실측의 경우, 측정 장비간 특성으로 인해 평균 공간 선량률 분포는 상이한 결과를 보였으나, 거리별 납 당량의 증가에 따라 지수함수분포로 공간 선량률이 감소되었다.

• Journal of Radiation Protection and Research
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• 제41권2호
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• pp.123-131
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• 2016

Background: With the increase in the number of particle accelerator facilities under either operation or construction, the accurate calculation using Monte Carlo codes become more important in the shielding design and radiation safety evaluation of accelerator facilities. Materials and Methods: The calculations with different physics models were applied in both of cases: using only physics model and using the mix and match method of MCNPX code. The issued conditions were the interactions of 600 MeV proton and $290MeV{\cdot}n^{-1}$ oxygen with a carbon target. Both of cross-section libraries, JENDL High Energy File 2007 (JENDL/HE-2007) and LA150, were tested in this calculation. In the case of oxygen ion interactions, the calculation results using LAQGSM physics model and JENDL/HE-2007 library were compared with D. Satoh's experimental data. Other Monte Carlo calculations using PHITS and FLUKA codes were also carried out for further benchmarking study. Results and Discussion: It was clearly found that the physics models, especially intra-nuclear cascade model, gave a great effect to determine proton-induced secondary neutron spectrum in MCNPX code. The variety of physics models related to heavy ion interactions did not make big difference on the secondary particle productions. Conclusion: The variations of secondary neutron spectra and particle transports depending on various physics models in MCNPX code were studied and the result of this study can be used for the shielding design and radiation safety evaluation.

• 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.

• Advances in materials Research
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• 제2권2호
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• pp.93-98
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• 2013

Objective of this study is to calculate gamma buildup factors for pointed and isotropic gamma sources in depleted uranium, uranium dioxide, natural uranium, tin, water and concrete using MCNP4C code. The thickness of the media ranges from 0.5 to 10 mean-free-path (mfp) and gamma energy ranges from 0.5 to 10 MeV. Owing to the outstanding accuracy of MCNP in calculation involving gamma interaction, results fairly match those reported previously. The maximum relative error is 2%.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2412-2420
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• 2024

The International Atomic Energy Agency (IAEA) mandates safeguards to ensure non-proliferation of nuclear materials. Among inspection techniques used to detect partial defects within spent nuclear fuel (SNF), gamma emission tomography (GET) has been reported to be reliable for detection of partial defects on a pin-by-pin level. Conventional GET, however, is limited by low detection efficiency due to the high density of nuclear fuel rods and self-absorption. This paper proposes a new type of GET named Spent Fuel Internal Tomography (SFIT), which can acquire sinograms at the guide tube. The proposed device consists of the housing, shielding, C-shaped collimator, reflector, and gadolinium aluminum gallium garnet (GAGG) scintillator. For accurate attenuation correction, the source-distinguishable range of the SFIT device was determined using MC simulation to the region away from the proposed device to the second layer. For enhanced inspection accuracy, a proposed specific source-discrimination algorithm was applied. With this, the SFIT device successfully distinguished all source locations. The comparison of images of the existing and proposed inspection methods showed that the proposed method, having successfully distinguished all sources, afforded a 150 % inspection accuracy improvement.