• 대한안전경영과학회지
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• 제17권1호
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• pp.103-109
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• 2015

In this study, we quantitatively analyzed the data by measuring the radiation shielding rate and uniformity in order to evaluate the performance of an Apron. In addition, storage conditions were also evaluated. The uniformity measurement was performed by evaluating the Apron DICOM images using a PACS program. The experiment was intended for 51 Aprons being used in three hospitals in the Daejeon area. The radiation shielding rate and uniformity were measured per lead equivalent for 0.25 mmPb, 0.35 mmPb, and 0.5 mmPb. As a result, the higher lead equivalents were, the greater differences in the non-uniformity between the top part and the bottom part became (p=0.020). In all hospitals, regarding the non-uniformity of four places in Aprons, all showed statistically significant differences (p<0.01). The average value of the transmitted radiation dose showed less difference (p=0.005) in the bottom right than in the upper right but was statistically significant. There have been no marks of manufacturing date or the date of purchase in the Apron.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2011년도 추계 학술논문 발표대회
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• pp.53-54
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• 2011

The facilities producing the nuclear energy chosen for resolving the recent global energy problem have been increasingly constructed, and hence more frequent durability tests on radiation shielding concrete are required due to NPP(Nuclear Power Plant) life extension and increase of radioactive waste repositories. Bulk dry density is one of the critical factors ensuring the durability and performance of the radiation shielding concrete because the design of the radiation shielding reinforced concrete structures for NPPs is based on the bulk dry density of the concrete. Bulk density of unconsolidated shielding concrete can be calculated utilizing a test assuring to satisfy the bulk dry density, or existing credible data set. This study evaluated correlation between bulk density and bulk dry density of the concrete used for Korean NPPs (y=1.0913X-0.2458) and developed a correlation expression considering standard deviation of bulk dry density (y=1.0913X-0.3358).

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.695-700
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• 2013

Thirty-eight radiation safety reports for brachytherapy equipment were evaluated to determine the current status of brachytherapy units in Korea and to assess how radiation oncology departments in Korea complete radiation safety reports. The following data was collected: radiation safety report publication year, brachytherapy unit manufacturer, type and activity of the source that was used, affiliation of the drafter, exposure rate constant, the treatment time used to calculate workload and the HVL values used to calculate shielding design goal values. A significant number of the reports (47.4%) included the personal information of the drafter. The treatment time estimates varied widely from 12 to 2,400 min/week. There was acceptable variation in the exposure rate constant values (ranging between 0.469 and 0.592 ($R{\cdot}m^2/Ci{\cdot}hr$), as well as in the HVLs of concrete, steel and lead for Iridium-192 sources that were used to calculate shielding design goal values. There is a need for standard guidelines for completing radiation safety reports that realistically reflect the current clinical situation of radiation oncology departments in Korea. The present study may be useful for formulating these guidelines.

• 한국수소및신에너지학회논문집
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• 제30권6호
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• pp.540-548
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• 2019

This study was conducted to develop insulation for solid oxide fuel cell (SOFC). The developed insulation is based on the lamination technology and the radiation shielding technology of the satellite insulation. The insulation material is consisting of insulation material for conduction resistance, spacer, and radiation shielding material. The experimental apparatus consisting vacuum bell jar, pump, heater and temperature recording device has developed to verify the performance of the insulation. The experimental values were used as reference data for the modeling development. In this paper, heat transfer is assumed to be one- dimensional phenomena for the prediction of insulation performance and internal temperature distribution in high temperature region of SOFC. The developed model was used to compare the performance difference of insulation types according to composition materials. The analysis result shows that the insulation including spacer and radiation shielding has better heat insulation performance than other cases. In this study, the thickness reduction effect of about 20% was shown compared to the insulation including only conductive material. It is noted that the radiant shielding material should be carefully selected for durability, because SOFC insulation should be used for a long time at high temperature.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.847-852
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• 1995

A shielding analysis for KSC-7, the shipping cask for transporting the 7 PWR spent fuel assemblies, has been carried out. Radiation source term has been calculated on spent fuel with burnup of 50,000 MWD/MTU and 1.5 years cooling time by ORIGEN2 code. The shielding calculation for the cask has been made by using MCNP4A code with continuous cross section data library from ENDF/B-V. As a result of neutron dose rate analysis, another shielding calculational model on spent fuel shipping cask was provided which is using the Monte Carlo method.

• 한국방사선학회논문지
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• 제15권5호
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• pp.697-704
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• 2021

3D 프린팅 FDM방식의 재료인 필라멘트 중 차폐성능을 지닌 필라멘트는 국내에 판매되지 않고 있으며 관련 연구도 미비하다. 이에 본 연구는 금속 입자가 함유된 필라멘트의 물성과 방사선의 차폐능력을 평가하여 3D 프린트를 이용한 방사선 차폐체 개발의 기초자료를 제공하고자 한다. 금속입자 강화재가 함유된 금속 필라멘트 5가지를 선정 후 ASTM의 평가방법을 이용하여 인장강도, 밀도, XRD, 무게측정 등 물성을 평가하고 방사선 차폐능력을 알아보기 위하여 한국산업표준의 방호용구류 시험방법에 따라서 방사선 차폐율을 구하였다. 인장강도는 PLA + SS가 가장 높았고 ABS + W가 가장 낮았으며 밀도는 ABS + W 가 3.13 g/cm3으로 가장 높게 나타났다. XRD결과 시편의 표면의 입자의 XRD peak 패턴이 각 입자 강화재 분말 금속의 패턴과 일치함을 확인 할 수 있어 프린트된 시편이 분말금속이 함유 되었음을 확인하였다. 3D 프린트 복합 필라멘트별 차폐효과는 ABS + W, ABS + Bi, PLA + SS, PLA + Cu, PLA + Al의 순서로 실효원자번호와 밀도에 비례하여 차폐율이 높게 나타났다. 본 연구에서는 강화재로 금속 분말이 함유된 금속입자 복합 필라멘트는 방사선의 차폐능력을 가지는 것이 확인되었으며 향후 방사선 차폐용 필라멘트의 사용가능성을 확인하였다.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2380-2387
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• 2023

Hyperspectral data and its ability to explore the minerals and their associated rocks have a remarkable application in mineral exploration and lithological characterization. The present study aims to explore the radiation shielding aspects of the iron ore in Kerala with the aid of the Hyperion hyperspectral dataset. The reflectance-spectra obtained from the laboratory conditions as well as from the image show various absorptions. The results from the spectra are validated with geochemical data and GPS points. The Monte Carlo simulation employed to evaluate the radiation shielding ability. Raising the oxygen ions caused a noteworthy decrease in the µ values of the studied rocks which is accompanied by an increase in Δ_0.5 and Δ_eq values. The Δ_0.5 and Δ_eq values increased by factors of approximately 77 % with raising the oxygen ions between 44.32 and 47.57 wt.%. The µ values varies with the oxygen concentrations, where the µ values decreased from 2.531 to 0.925 cm^-1 (at 0.059 MeV), from 0.381to 0.215 cm^-1 (at 0.662 MeV), and from 0.279 to 0.158 cm^-1 (at 1.25 MeV) with raising the oxygen ions from 44.32 to 47.43 wt.%.

• 한국방사선학회논문지
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• 제17권5호
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• pp.641-649
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• 2023

본 연구에서는 3D 프린팅 기술을 활용하여 제작한 차폐체의 성능을 비교 분석하여, 고에너지 전자선 치료 시 차폐체로서의 적용 가능성에 대해 알아보고자 한다. 고에너지 전자선에 대한 3D 프린팅 재료의 차폐성능 평가를 위해 실측과 몬테카를로 기반의 모의실험을 수행하였다. 첫 번째, 모의실험에 대한 신뢰성 확보를 위해 IAEA의 TRS-398 권고를 참조하여 선원항 평가를 수행하였다. 두 번째, PLA+W (93%) 재료에 대한 차폐 성능 분석을 위해 3D 프린터를 이용하여 시편을 제작하였고, 전자선 에너지에 따른 두께별 차폐율을 평가하였다. 세 번째, PLA+W (93%)와 기존 차폐체 간 차폐 성능 비교 분석을 통해 전자선 치료 시 필요한 차폐 두께를 산정하였다. 연구 결과, 첫 번째, 실측과 모의실험을 통한 선원항 평가 결과, 1% 이내의 오차로 TRS-398 권고를 만족하여 모의실험에 대한 신뢰성을 확보하였다. 두 번째, PLA+W (93%)에 대한 차폐 성능 분석 결과, 6 MeV 전자선은 3.12 mm에서 95% 이상의 차폐율을 나타냈고, 15 MeV 전자선은 10 mm 두께에서 90% 이상의 차폐율을 나타내었다. 세 번째, 모의실험을 통해 PLA+W (93%) 재료와 기존 차폐체 간 비교 분석을 통해 동일 두께 내에서 텅스텐, 납, 구리, PLA+W (93%), 알루미늄 순서로 차폐율이 높은 결과를 나타내었으며, 6 MeV 전자선은 5 mm 이상, 15 MeV 전자선은 10 mm 이상 두께에서 거의 유사한 차폐율을 나타내었다. 향후 본 연구를 통해 고에너지 전자선 치료 시 PLA+W (93%) 재료를 이용한 환자의 맞춤형 차폐체 제작을 위한 기초자료로서 활용될 수 있을 것으로 판단된다.

• 대한방사선기술학회지:방사선기술과학
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• 제41권2호
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• pp.97-102
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• 2018

The purpose of this study is to provide the basic data for reducing unnecessary radiation dose to the abdomen and fetus of pregnant women by presenting proper height of shielding protector for efficient abdominal shielding in chest PA examination of Korean pregnant women. The subjects of this study were 288 persons who were eligible for this study among 798 pregnant women who had chest PA examination from January 1, 2015 to December 31, 2016 Retrospective study was performed. Measurements was performed from the apex of the right and left lungs to costophrenic angle of the right and left lungs and to the lowest costophrenic angle among the right and left lungs at the top of the image(this line called Joo's line in this study). The mean of the right and left lung height of pregnant women were 259.09 mm and 263.57 mm, respectively. Also, the average height of the Joo's line designed by the researcher for proper abdominal radiation protection was 322.15 mm. For proper and efficient abdominal radiation protection for pregnant women, it is necessary to adjust the shielding according to the height of the pregnant woman. It is appropriate that the height of the shielding protector should be adjusted so that the upper part of the shield is located at 342.30 mm below from upper part of the detector.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1784-1791
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• 2020

The 14 MeV neutron generator facility is being developed by the Institute for Plasma Research India to conduct the lab scale experiments related to Indian breeding blanket system for ITER and DEMO. It will also be utilized for material testing, shielding experiments and development of fusion diagnostics. Occupational radiation exposure control is necessary for the all kind of nuclear facilities to get the operational licensing from governing authorities and nuclear regulatory bodies. In the same way, the radiation exposure for the 14 MeV neutron generator facility at the occupational worker area and accessible zones for general workers should be under the permissible limit of AERB India. The generator is designed for the yield of 10¹² n/s. The shielding assessment has been made to estimate the radiation dose during the operational time of the neutron generator. The facility has many utilities and constraints like ventilation ducts, accessible doors, accessibility of neutron generator components and to conduct the experiments which make the shielding assessment challenging to provide proper safety for occupational workers and the general public. The neutron and gamma dose rates have been estimated using the MCNP radiation transport code and ENDF -VII nuclear data libraries. The ICRP-74 fluence to dose conversion coefficients has been used for the assessment. The annual radiation exposure has been assessed by considering 500 h per year operational time. The provision of local shield near to neutron generator has been also evaluated to reduce the annual radiation doses. The comprehensive results of radiation shielding capability of neutron generator building and local shield design have been presented in the paper along with detailed maps of radiation field.