• Nuclear Engineering and Technology
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• 제47권3호
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• pp.380-387
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• 2015

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as $^{235}U$, $^{239}Pu$, $^{241}Pu$, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux ($>10^{12}n/cm^2{\cdot}s$) neutron source comprised of a high-energy (30 MeV)/high-current (~2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (< $0.06{\mu}Sv/h$), a few shielding materials [high-density polyethylene (HDPE)eBorax, $B_4C$, and $Li_2CO_3$] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.

• 한국의학물리학회지:의학물리
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• 제33권4호
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• pp.164-171
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• 2022

The number of facilities using radiation generators increases and related regulations are strengthened, the establishment of a shielding management and evaluation technology has become important. The characteristics of the radiation generator used in previous report differ from those of currently available high-frequency radiation generators. This study aimed to manufacture lead, iron, and concrete shielding materials for the re-verification of half-value layers, tenth-value layers, and attenuation curve. For a comparison of attenuation ratio, iron, lead, and concrete shields were manufactured in this study. The initial dose was measured without shielding materials, and doses measured under different types and thicknesses of shielding material were compared with the initial dose to calculate the transmission rate on 50-300 kVp X-ray. All the three shielding materials showed a tendency to require greater shielding thickness for higher energy. The attenuation graph showed an exponential shape as the thickness decreased and a straight line as the thickness increased. The difference between the measurement results and the previous study, except in extrapolated parts, may be due to the differences in the radiation generation characteristics between the generators used in the two studies. The attenuated graph measured in this study better reflects the characteristics of current radiation generators, which would be more effective for shield designing.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1210-1217
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• 2023

The assessment of radiation fields in the lithium loop pipes and dump tank during the operation were performed for International Fusion Materials Irradiation Facility - DEMO-Oriented NEutron Source (IFMIF-DONES) in order to obtain the radiation dose-rate maps in the component surroundings. Variance reduction techniques such as weight window mesh (produced with the ADVANTG code) were applied to bring the statistical uncertainty down to a reasonable level. The biological dose was given in the study, and potential shielding optimization is suggested and more thoroughly evaluated. The MCNP Monte Carlo was used to simulate a gamma particle transport for radiation shielding purposes for the current Li Systems' design. In addition, the shielding efficiency was identified for the Impurity Control System components and the dump tank. The analysis reported in this paper takes into account the radiation decay source from and activated corrosion products (ACPs), which is created by d-Li interaction. As a consequence, the radiation (resulting from ACPs and Be-7) shielding calculations have been carried out for safety considerations.

• Radiation Oncology Journal
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• 제13권1호
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• pp.101-104
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• 1995

High Dose Rate Remote Afterloading system was installed at Wonkwang University Hospital in January 1994. In this report, the calibration of a Gammamed 12-i High Dose Rate Remote Afterloading system and the radiation survey around the facility after design and construct a shieding room are discussed. The radiation survey of the facility indicates that the use of ordinary concrete shielding of existing room will provide adequate shielding. Also, the methodologies for performing source calibration are presented.

• Journal of Radiation Protection and Research
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• 제48권4호
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• pp.184-196
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• 2023

Background: Radiation protection is crucial in various fields due to the harmful effects of radiation. Shielding is used to reduce radiation exposure, but gamma radiation poses challenges due to its high energy and penetration capabilities. Materials and Methods: This work investigates the radiation shielding properties of polyvinylidene fluoride (PVDF) samples containing different weight fraction of tungsten carbide (WC), tungsten trioxide (WO₃), and tungsten disulfide (WS₂). Parameters such as the mass attenuation coefficient (MAC), half-value layer (HVL), mean free path (MFP), effective atomic number (Z_eff), and macroscopic effective removal cross-section for fast neutrons (Σ_R) were calculated using the Phy-X/PSD software. EpiXS simulations were conducted for MAC validation. Results and Discussion: Increasing the weight fraction of the additives resulted in higher MAC values, indicating improved radiation shielding. PVDF-xWC showed the highest percentage increase in MAC values. MFP results indicated that PVDF-0.20WC has the lowest values, suggesting superior shielding properties compared to PVDF-0.20WO₃ and PVDF-0.20WS₂. PVDF-0.20WC also exhibited the highest Z_eff values, while PVDF-0.20WS₂ showed a slightly higher increase in Z_eff at energies of 0.662 and 1.333 MeV. PVDF-0.20WC has demonstrated the highest Σ_R value, indicating effective shielding against fast neutrons, while PVDF-0.20WS₂ had the lowest Σ_R value. The Monte Carlo N-Particle Transport version 5 (MCNP5) simulations showed that PVDF-xWC attenuates gamma radiation more than pure PVDF, significantly decreasing the dose equivalent rate. Conclusion: Overall, this research provides insights into the radiation shielding properties of PVDF mixtures, with PVDF-xWC showing the most promising results.

• 한국방사선학회논문지
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• 제18권1호
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• pp.11-20
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• 2024

최근 인터벤션의 중요성이 증가한 만큼 시술을 수행하는 의료진의 건강에 대한 관심이 높아지고 있다. 기존의 방사선 차폐 기구는 시술자의 동선을 제한하고 감염의 위험으로 인해 적절하게 사용되지 못했으며 시술자의 생식선과 나아가 시술실 전체 영역의 적절한 방사선 차폐가 이루어지지 못했다. 시술 시 사용되는 팔꿈치 지지대에 차폐체(bismuth)를 부착하여 보조 차폐 기구를 제작하였고 방사선 차폐 효과를 측정하였다. 측정 결과, 평균 공간 선량률이 약 64.8% 감소하였으며 독립표본 T검정 분석 결과 유의확률 이하(p<0.05)로 통계학적으로 유의미하게 나타났다. 보조 차폐 기구의 사용은 시술자의 생식선 차폐 및 시술실 전체 영역의 방사선 공간 선량률을 감소시킬 수 있는 효과적인 차폐 방법으로 사료된다.

• 한국방사선학회논문지
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• 제16권3호
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• pp.233-240
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• 2022

용융적층 방식의 필라멘트에 대한 방사선의 차폐유무의 관한 연구가 최근 연구되어지기 시작하였지만 차폐능력을 가진 필라멘트는 국내에 판매되지 않고 있으며 관련 연구도 미비하다. 이에 본 연구는 고밀도 폴리에틸렌을 기지재로 하고 강화재로 비스무트를 선정하여 복합 필라멘트를 제작한 후 차폐능력을 평가하고 3D 프린트를 이용한 방사선 차폐 복합물질 개발의 기초자료를 제공하고자 한다. 고밀도 폴리에틸렌에 실효 원자번호가 83인 비스무트를 혼합하였고 비스무트의 함유량을 20 wt%, 30 wt%, 40 wt%로 조절하여 필라멘트를 제작하였다. 제작된 필라멘트는 ASTM의 평가방법을 이용하여 물성 및 차폐능력을 평가하였다. 비스무트 함유량이 증가할수록 밀도, 무게, 인장강도는 증가하였고 차폐능력이 우수해짐을 확인 할 수 있었다. 방사선 차폐능력 평가 결과 HDPE(80%) + Bi(20%)의 경우 60 kV일 때 82%의 차폐율을 보였으며 비스무트 함유량이 40% 일 때는 최대 94.57%이상의 차폐율을 나타내는 것을 확인하였다. 본 연구에서는 HDPE + Bi 필라멘트를 사용하면 기존에 연구되어진 금속 입자 함유 필라멘트들보다 가볍고 방사선을 차폐할 수 있는 방사선 차폐체 제작이 가능하다는 것을 확인하였고 의료 및 방사선 산업에 있어 방사선 차폐 복합물질로서의 사용가능성을 확인하였다.

• Journal of Radiation Protection and Research
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• 제32권4호
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• pp.190-193
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• 2007

The objective of this study is to recommend the radiation protection design parameters from the shielding point of view for concrete wall between the decay tank room and the primary pump room in TRIGA Mark-II Research Reactor Facility. The shield design for this concrete wall has been performed with the help of Point-kernel Shielding Code Micro-Shield 5.05 and this design was also validated based on the measured dose rate values with Radiation Survey Meter (G-M Counter) considering the ICRP-60 (1990) recommendations for occupational dose rate limit ($10{\mu}Sv/hr$). The recommended shield design parameters are: (i) thickness of 114.3 cm Ilmenite-Magnetite Concrete (IMC) or 129.54 cm Ordinary Reinforced Concrete (ORC) for concrete wall A (ii) thickness of 66.04 cm Ilmenite-Magnetite Concrete (IMC) or 78.74 cm Ordinary Reinforced Concrete (ORC) for concrete wall B and (iii) door thickness of 3.175 cm Mild Steel (MS) on the entrance of decay tank room. In shielding efficiency analysis, the use of I-M concrete in the design of this concrete wall shows that it reduced the dose rate by a factor of at least 3.52 times approximately compared to ordinary reinforced concrete.

• 한국방사선학회논문지
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• 제15권6호
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• pp.813-819
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• 2021

본 연구에서는 방사성 불소의 인체 내 주입 시 방사선 방호목적으로 사용되는 L-블럭형 방호장비의 차폐율과 주변공간의 선량분포를 몬테칼로 방법을 이용해 계산하였다. L-블럭형 차폐장치의 몸체 및 윈도우 부위의 차폐율은 99.99 %였다. 1 m거리에서 계산한 선량분포는 XZ평면의 135°, 45°, 225°, 315°, 180°에서 상대적으로 높게 나타났고, 0°, 90°, 270°에서는 매우 낮게 계산되었다. YZ평면에서는 135°, 180°, 225°에서 상대적으로 높게 나타났고, 나머지 각도에서는 매우 낮게 계산되었다. AZ와 BZ 평면에서도 YZ평면과 유사한 결과를 나타냈다. 또한 선원의 수평방향과 선원의 상방 45°방향의 선량분포를 통해 225°~315°범위에서 차폐율이 가장 우수함을 확인하였다. 이와 같은 결과가 방사선 작업 종사자들의 방사선 방호에 필요한 기초자료로 활용되기를 기대한다.

• International Journal of Contents
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• 제14권4호
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• pp.65-69
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• 2018

The objective of this study was to conduct a radiation shielding analysis for the facility equipped with radiation generator. The analysis was carried out in two aspects. First, from the aspect of the effect caused by primary and leakage radiation. Second, effect of scattered radiation was evaluated by applying a simple calculation method based on a scattering rate concept since effect of scattered radiation is significantly important at maze entrance of the radiation facility. The calculated results obtained using the simple method were compared to the results calculated using Geant4 code and the measured values. The results calculated by the suggested method indicate that slight error exists in a radiation shielding analysis done at the maze entrance comparing to other two results, while the results evaluated at the outside of the maze entrance door are relatively consistent with other values.