• 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.

• 한국정보통신학회:학술대회논문집
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• 한국정보통신학회 2013년도 춘계학술대회
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• pp.1031-1033
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• 2013

후쿠시마와 같은 방사선 누출 사고 시 방사선원의 위치를 찾는 일은 방사선 방호 뿐만 아니라 원전 사고의 조속하고 안전한 처리를 위해서도 중요하다. 방사선원의 3차원 위치 탐지는 기존에 방사선 탐지기의 2차원적 방사선 위치 탐지기능에 방사선원의 거리정보까지 추가 제공할 수 있어 방사선 오염원의 제거 및 제염작업에 결정적 역할을 할 수 있다. 본 연구에서는 반도체 센서에 기반한 듀얼(Dual) 방사선 탐지기를 이용한 방사선원 3차원 가시장치 개발 연구의 일환으로 방사선 센서부의 효율적 차폐체 구조설계에 관한 결과를 논하였다. 고하중의 텅스텐 또는 납 차폐체를 MCNP기반으로 최적구조로 설계함으로써 경량의 고효율 방사선원 위치탐지기 구현을 시도하였고, 이를 위해 차폐체의 구조와 두께, 그리고 콜리메이터에 형상의 다양한 변수모델에 대한 방사선 차폐시뮬레이션을 수행하였다. 본 연구의 결과는 향후 실리콘 센서기반의 소형 경량의 3차원 방사선원 탐지 및 가시화 연구에 활용될 예정이다.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.178-184
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• 2020

The traditional methods for radiation shield design always only focus on either the structure or the components of the shields rather than both of them at the same time, which largely affects the shielding performance of the facilities, so in this paper, a novel method for designing the structure and components of shields simultaneously is put forward to enhance the shielding ability. The method is developed by using the genetic algorithm (GA) and the MCNP software. In the research, six types of shielding materials with different combinations of elements such as polyethylene (PE), lead (Pb) and Boron compounds are applied to the radiation shield design, and the performance of each material is analyzed and compared. Then two typical materials are selected based on the experiment result of the six samples, which are later verified by the Compact Accelerator Neutron Source (CANS) facility. By using this method, the optimal result can be reached rapidly, and since the design progress is semi-automatic for most procedures are completed by computer, the method saves time and improves accuracy.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2013년도 춘계 학술논문 발표대회
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• pp.35-36
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• 2013

Electric arc furnace oxidizing slag from massively produced steel slag has been used in road bases and subbases, hot mix asphalt, and landfill. Electric arc furnace oxidizing slag contains iron (15%~30%) and has a high density of 3.0~3.7 ton/m3. Depending on the type and amount of concrete aggregates, the radiation-shielding characteristics can vary. Therefore, aggregates of electric arc furnace oxidizing slag can be considered for the production of radiation-shielding concrete. The experimental design of this study is experiments on Compressive strength experiments, X-ray irradiation experiments, and experiments related to the unit volume weight were carried out on hardened concrete. This experiment compared the performance evaluation of radiation shielding of concrete using electric arc furnace oxidizing slag.

• 한국방사선학회논문지
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• 제16권5호
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• pp.527-536
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• 2022

본 연구는 우주방사선 차폐물질 설계를 위한 선행연구 차원에서 우주방사선에 대한 물질별 방사선 차폐특성을 분석하였다. 특히 EMP 및 방사선 차폐에 효과가 있다고 알려진 경량 연자성 복합소재에 대한 우주방사선 차폐물질 활용 가능성을 확인하고자 하였다. 이를 위해 Monte Carlo N-Particle(MCNP) 모델링 기법과 열중성자 차폐실험을 수행하였으며, MCNP의 우주방사선 모델인 Skymap.dat를 활용하였다. 연구결과 폴리에틸렌, 붕소폴리에틸렌, 탄소나노튜브 등 탄소와 수소를 함유한 물질의 경우 증발 중성자 에너지 영역 대 이하의 중성자 감소에 효과적인 것으로 나타났으며 SS316, 경량 연자성 물질 등 철을 함유한 물질은 캐스케이드 중성자 차폐성능이 뛰어난 것을 확인할 수 있었다. 특히 경량 연자성 물질의 경우 붕소를 함유하고 있어 저속중성자 영역의 중성자 감소에도 효과적인 것으로 나타났으며, 향후 탄소 및 수소 등 탄성산란 물질을 보강한다면 우주방사선 중성자 전 영역에서 유의미한 차폐효과를 보여줄 것으로 기대된다.

• 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.

• 한국의학물리학회지:의학물리
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• 제32권4호
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• pp.83-91
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• 2021

Various types of high-precision radiotherapy, such as intensity-modulated radiation therapy (IMRT), tomotherapy (Tomo), and stereotactic body radiation therapy have been available since 1997. After being covered by insurance in 2015, the number of IMRT cases rapidly increased 18-fold from 2011 to 2018 in Korea. IMRT, which uses a high-beam irradiation monitor unit, requires higher shielding conditions than conventional radiation treatments. However, to date, research on the shielding of facilities using IMRT and the current understanding of its status are insufficient, and detailed safety regulation procedures have not been established. This study investigated the recommended criteria for the shielding evaluation of facilities using medical linear accelerators (LINACs), including 1) the current status of safety management regulations and systems in domestic and international facilities using medical LINACs and 2) the current status of the recommended standards for safety management in domestic and international facilities using medical LINACs. It is necessary to develop and introduce a safety management system for facilities using LINACs for clinical applications that is suitable for the domestic medical environment and corresponds to the safety management systems for LINACs used overseas.

• 대한방사선치료학회지
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• 제2권1호
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• pp.31-40
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• 1987

In the case of designing a high dose rate remote controlled afterloading treatment room with existing hospital facilities. We must construct the effective protective barriers so as to reduce the primary and scattered radiation up to the maximum permissible dose level. It is difficult to reinforce the barrier thickness of the shielding requirements because of the limited space and the problem of the existing building structure at the surrounding area. Therefore we can reduce the intensity of primary radiation to the required degree at the location of interest with installing the appropriate I shaped Pb barriers between the radiation source and the shielding wall of the concrete. As a result, it was possible to reduce the intensity of the primary radiation below the M.P.D level by using additional Pb barriers instead of increasing thickness of concrete wall.

• 한국의학물리학회지:의학물리
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• 제29권4호
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• pp.115-122
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• 2018

Although the current internationally recommended standard for the use factor (U) applied to CyberKnife is 0.05 (5%), the CyberKnife shielding standard is applied more stringently. This study, based on clinical data, was aimed at examining the appropriateness of existing shielding guidelines. Sixty patients treated with G4 CyberKnife were selected. The patients were divided into two groups, according to whether they underwent skull or spine tracking. Based on the results, the use factors for each wall ranged from 0.028 (2.8%) to 0.031 (3.1%) for the intracranial treatment and 0.020 (2.0%) to 0.022 (2.2%) for the body treatment. Excessive barrier thickness resulted in inefficient use of space and higher cost to the institutions. Furthermore, because the use factor is influenced by the position of the robot, the use factor determined based on the clinical data of this study would facilitate more reasonable treatment room design.

• Journal of Radiation Protection and Research
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• 제47권1호
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• pp.50-57
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• 2022

Background: There are several proton therapy facilities in operation or planned in Taiwan, and these facilities are anticipated to not only treat cancer but also provide beam services to the industry or academia. The simplified approach based on the Monte Carlo-based data sets (source terms and attenuation lengths) with the point-source line-of-sight approximation is friendly in the design stage of the proton therapy facilities because it is intuitive and easy to use. The purpose of this study is to expand the Monte Carlo-based data sets to allow the simplified approach to cover the application of proton beams more widely. Materials and Methods: In this work, the MCNP6 Monte Carlo code was used in three simulations to achieve the purpose, including the neutron yield calculation, Monte Carlo-based data sets generation, and dose assessment in simple cases to demonstrate the effectiveness of the generated data sets. Results and Discussion: The consistent comparison of the simplified approach and Monte Carlo simulation results show the effectiveness and advantage of applying the data set to a quick shielding design and conservative dose assessment for proton therapy facilities. Conclusion: This study has expanded the existing Monte Carlo-based data set to allow the simplified approach method to be used for dose assessment or shielding design for beam services in proton therapy facilities. It should be noted that the default model of the MCNP6 is no longer the Bertini model but the CEM (cascade-exciton model), therefore, the results of the simplified approach will be more conservative when it was used to do the double confirmation of the final shielding design.