• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2613-2620
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• 2023

Excellent thermal neutron absorption performance of boron expands the potential use of boron rich slag to prepare epoxy resin matrix nuclear shielding composites. However, shielding attenuation behaviors and mechanism of the composites against gamma rays are unclear. Based on the radiation protection theory, Phy-X/PSD, XCOM, and ⁶⁰Co gamma ray source were integrated to obtain the shielding parameters of boron rich slag/epoxy resin composites at 0.015-15 MeV, which include mass attenuation coefficient (µ_t), linear attenuation coefficient (µ), half value thickness layer (HVL), electron density (N_eff), effective atomic number (Z_eff), exposure buildup factor (EBF) and exposure absorption buildup factor (EABF).µ_t, µ, HVL, N_eff, Z_eff, EBF and EABF are 0.02-7 cm²/g, 0.04-17 cm^-1, 0.045-20 cm, 5-14, 3 × 10²³-8 × 10²³ electron/g, 0-2000, and 0-3500. Shielding performance is BS4, BS3, BS3, BS1 in descending order, but worse than ordinary concrete. µ and HVL of BS1-BS4 for ⁶⁰Co gamma ray is 0.095-0.110 cm^-1 and 6.3-7.2 cm. Shielding mechanism is main interactions for attenuation gamma ray by BS1-BS4 are elements with higher content or higher atomic number via Photoelectric Absorption at low energy range, and elements with higher content via Compton Scattering and Pair Production in Nuclear Field at middle and higher energy range.

• 방사성폐기물학회지
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• 제16권2호
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• pp.243-260
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• 2018

사용후핵연료 건식저장 시스템과 관련하여 고온 및 방사선으로 인한 콘크리트 손상과 열화특성에 대해 포괄적으로 문헌분석을 수행하였다. 고온에 의한 장기열화를 방지하기 위한 콘크리트의 임계온도는 일반적으로 $95^{\circ}C$이며, 온도경사는 콘크리트 균열방지를 위해 $60^{\circ}C$ 이하가 되도록 설정하고 있다. 열화정도는 노출온도와 노출시간에 비례하여 증가하는 경향을 나타내며, 압축강도에 비해 인장강도가 고온에 보다 민감한 특성을 보인다. 한편 방사선의 에너지가 $10^{10}MeV{\cdot}cm^{-2}{\cdot}s^{-1}$ 이하일 경우에는 핵반응으로 인한 가열을 무시할 수 있다. 하지만 콘크리트가 $10^{19}n{\cdot}cm^{-2}$ 이상의 중성자에 혹은 $10^{10}$ rad를 초과하는 감마선량에 노출된다면 콘크리트의 역학적 물성이 점차 감소하는 경향을 보이며, 그 손상정도는 콘크리트 구성재료의 특성에 의존적이다. 콘크리트에 대한 방사선 조사시 재료의 역학적 물성변화는 주로 온도상승으로 인한 콘크리트 내부 함수량의 변화 및 재료간의 열적물성 차이로 인한 체적증가와 균열발생으로 발생한다. 따라서 건식저장과 관련된 기술의 조속한 확보 및 인 허가를 위해서는 그 간의 선행연구 결과를 최대한 활용할 필요가 있으며, 본 연구결과는 향후 사용후핵연료 건식저장 콘크리트 캐스크 관련 국내 자체기술 개발에 중요한 기초자료로 활용될 수 있을 것이다.

• Nuclear Engineering and Technology
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• 제3권2호
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• pp.65-75
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• 1971

국내에서 산출되는 각종 광물골재를 사용하여 방사선 차폐용 중차폐 콩크리트를 제조하고 감마선에 대한 차폐 효과를 실험한 결과 최적하다고 판단된 자철광 중차폐 콩크리트를 대상으로 60Co 감마선의 Broad beam을 사용하여 방사선 차폐 효과를 측정하였다. 본 실험을 통하여 실험적으로 차폐체내의 방사선의 감쇄곡선으로부터 차폐 체 두께의 변화에 따르는 방사선 투과율과의 상호관계에 관한 수식을 다음과 같이 유도해냈다. I (x) = I (ο) exp(-$\mu$X) exp(1.03$\times$$10^{-1}$X-3.38$\times$$10^{-3}$X$^2$＋5.29$\times$$10^{-5}$X$^3$) X< 20 cm 때, I (x) =I (ο) exp(-$\mu$X) exp(4.66$\times$$10^{-2}$ X＋2.12$\times$$10^{-1}$) X＞20 cm 때. 이와같이 얻은 결과식에서 오른쪽 첫번째항은 최초 감마선의 감쇄를 표시하고 그 다음항은 차폐체 내에서의 감마선 재생계수를 나타낸다. 이 실험에 첨가하여 차폐체의 실제 설계에 입각한 입방형 자철광 구조체 (두께 8 cm, 내부공간 40$\times$40$\times$40cm)에 대한 차폐효과를 측정한 결과 평판 차폐체를 사용할 때 보다 투과 방사선이 증가됨을 알았다.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.547-556
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• 2014

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

• 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%.

• 대한방사선기술학회지:방사선기술과학
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• 제41권6호
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• pp.581-586
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• 2018

This study was conducted to analyze factors Influencing Protective Behavior against Radiation Exposure using questionnaires for 231 radiological technologists working in Computed Tomography(CT) examination room with high radiation dose in diagnostic radiology field. Statistical analysis of the collected data revealed that the reasons for partially shielding the examination part in the CT scan were the lack of protective equipment, securing of radiation justification, being annoying and maybe not being harm to adults in order. It was also revealed that the variables influencing the protective behavior were protective behavior against radiation harm, self-efficacy, protective environment, organization culture, protective knowledge and protective instrument in order. The higher the radiological protective environment(${\beta}=0.245$) and the lower the radiological protective knowledge(${\beta}=-0.034$), the more influential the protective behavior against radiation harm was. In this study, it was shown that non examination parts were not shielded in the CT scan. Therefore, it is necessary to improve the level of protective environment, to cultivate knowledge to improve the protective behavior against radiation harm and to have an intervention strategy for concrete action.

• 한국콘텐츠학회논문지
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• 제15권7호
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• pp.282-288
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• 2015

몬테카를로 기법을 기반으로 한 모의실험을 통해 방사성 핵종별 주사기 차폐기구의 재질 및 두께에 대한 차폐분석 결과, 텅스텐, 납, 비스무스와 같이 상대적으로 원자번호가 높은 재질의 경우 거의 모든 핵종에서 가장 높은 차폐효과를 보였다. 그러나 $^{18}F$, $^{67}Ga$, $^{111}In$ 선원의 경우, 차폐두께가 낮은 영역에서 저 원자번호 재질보다 더 높은 에너지를 나타냈으나, 이후 증가된 차폐두께에서는 투과되어 도달하는 감마선이 감소하여 더 낮은 에너지 분포를 나타냈다. 그 외 상대적으로 원자번호가 낮은 재질의 경우 구리, 철, 스테인리스강, 황산바륨의 순서로 에너지가 낮은 분포를 나타냈고, 알루미늄, 플라스틱, 콘크리트, 물의 경우 핵종별로 각기 다른 양상을 나타냈으며, 상대적으로 투과된 감마선의 증가로 전체적으로 높은 에너지 분포를 보여 차폐효과가 떨어지는 것으로 나타냈다.

• 한국산학기술학회논문지
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• 제11권3호
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• pp.1035-1039
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• 2010

고에너지 X-ray를 투시 방사선원으로 사용한 컨테이너 화물검색시스템에서 생성되는 광중성자에 의해 주변 콘크리트 차폐벽에서 발생되는 방사화생성물을 평가하였다. 몬테카를로 전산해석 코드인 MCNPX2.5.0을 사용하였으며, 참조시스템은 국내 주요 항만에 설치된 9MeV X-ray 고정식 양방향 컨테이너 화물검색시스템이다. 9MeV X-ray 조사에 따라 생성되는 광중성자의 (n,$\gamma$) 반응에 의한 방사화생성물 재고량을 계산하고 이에 따라 야기되는 방사선 피폭선량을 계산하였다.

• Nuclear Engineering and Technology
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• 제46권1호
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• pp.73-80
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• 2014

A concrete cask is an option for spent nuclear fuel interim storage. A concrete cask usually consists of a metallic canister which confines the spent nuclear fuel assemblies and a concrete overpack. When the overpack undergoes a missile impact, which might be caused by a tornado or an aircraft crash, it should sustain an acceptable level of structural integrity so that its radiation shielding capability and the retrievability of the canister are maintained. A missile impact against a concrete overpack produces two damage modes, local damage and global damage. In conventional approaches [1], those two damage modes are decoupled and evaluated separately. The local damage of concrete is usually evaluated by empirical formulas, while the global damage is evaluated by finite element analysis. However, this decoupled approach may lead to a very conservative estimation of both damages. In this research, finite element analysis with material failure models and element erosion is applied to the evaluation of local and global damage of concrete overpacks under high speed missile impacts. Two types of concrete overpacks with different configurations are considered. The numerical simulation results are compared with test results, and it is shown that the finite element analysis predicts both local and global damage qualitatively well, but the quantitative accuracy of the results are highly dependent on the fine-tuning of material and failure parameters.

• 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.