• 대한방사선방어학회:학술대회논문집
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• 대한방사선방어학회 2009년도 추계 학술발표
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• pp.200-201
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• 2009

• 대한방사선방어학회:학술대회논문집
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• 대한방사선방어학회 2009년도 추계 학술발표
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• pp.80-81
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• 2009

• 방사선산업학회지
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• 제17권1호
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• pp.93-100
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• 2023

This study analyzes the neutron shielding performance of Soft Magnetic Material and proposes a military application. In general, the military protection facility has been constructed with thick concrete, so Soft Magnetic Material, consisting of boron, was used with concrete in this study. To do so, Monte-Carlo N-Particle (MCNP) was applied to simulate the Watt-fission neutron spectrum of ²³⁵U and ²³⁹Pu. As a result, a configuration of polyethylene and Soft Magnetic Material is evaluated about four times better than borated polyethylene concerning the atomic weight of boron inside each shielding material. Also, when a nuclear weapon explosion is simulated in MCNP, 1 mm of Soft Magnetic Material with 20 cm of concrete shows about 55% more additional neutron shielding performance compared to when Soft Magnetic Material is not used. In this work, the neutron shielding performance of Soft Magnetic Material could be identified and Soft Magnetic Material would be useful for neutron shielding if applicable to concrete structure.

• 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.1817-1825
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• 2020

Background: Cosmic ray-induced particles can lead to failure of semiconductors packaged for export during air transport. This work performed MCNP 6.2 simulations to optimize shielding against neutrons and protons induced by cosmic radiation Methods and materials: The energy spectra of protons and neutrons by incident angle at the flight altitude were determined using atmospheric cuboid model. Various candidates for the shielding materials and the geometry of the Unit Load Device Container were evaluated to determine the conditions that allow optimal shielding at all sides of the container. Results: It was found that neutrons and protons, at the flight altitude, generally travel with a downward trajectory especially for the particles with high energy. This indicated that the largest number of particles struck the top of the container. Furthermore, the simulation results showed that, among the materials tested, borated polyethylene and stainless steel were the most optimal shielding materials. The optimal shielding structure was also determined with the weight limit of the container in consideration. Conclusions: Under the determined optimal shielding conditions, a significantly reduced number of neutrons and protons reach the contents inside the container, which ultimately reduces the possibility of semiconductor failure during air transport.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.277-284
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• 2017

This paper presents the establishment and characterization of a neutron calibration field using a bare $^{252}Cf$ source of low neutron source strength in Vietnam. The characterization of the field in terms of neutron flux spectra and neutron ambient dose equivalent rates were performed by Monte Carlo simulations using the MCNP5 code. The anisotropy effect of the source was also investigated. The neutron ambient dose equivalent rates at three reference distances of 75, 125, and 150 cm from the source were calculated and compared with the measurements using the Aloka TPS-451C neutron survey meters. The discrepancy between the calculated and measured values is found to be about 10%. To separate the scattered and the direct components from the total neutron flux spectra, an in-house shadow cone of 10% borated polyethylene was used. The shielding efficiency of the shadow cone was estimated using the MCNP5 code. The results confirmed that the shielding efficiency of the shadow cone is acceptable.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.633-646
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• 2020

Imaging and scattering techniques using thermal neutrons allow to analyze complex specimens in scientific and industrial researches. Owing to this advantage, there have been a considerable demand for neutron facilities in the industrial sector. Among neutron sources, an accelerator driven compact neutron source is the only one that can satisfy the various requirements-construction budget, facility size, and required neutron flux-of industrial applications. In this paper, a target, moderator, and reflector (TMR) system for low-energy proton-accelerator driven compact thermal neutron source was designed via Monte Carlo simulations. For 10-30 MeV proton beams, the optimal conditions of the beryllium target were determined by considering the neutron yield and the blistering of the target. For a non-borated polyethylene moderator, the neutronic properties were verified based on its thickness. For a reflector, three candidates-light water, beryllium, and graphite-were considered as reflector materials, and the optimal conditions were identified. The results verified that the neutronic intensity varied in the order beryllium > light water > graphite, the compacter size in the order light water < beryllium < graphite and the shorter emission time in the order graphite < light water < beryllium. The performance of the designed TMR system was compared with that of existing facilities and were laid between performance of existing facilities.

• Journal of Radiation Protection and Research
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• 제45권3호
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• pp.130-141
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• 2020

Background: Concrete activation in cyclotron vaults is a major concern associated with their decommissioning because a considerable amount of activated concrete is generated by secondary neutrons during the operation of cyclotrons. Reducing the amount of activated concrete is important because of the high cost associated with radioactive waste management. This study aims to investigate the capability of the neutron absorbing materials to reduce concrete activation. Materials and Methods: The Particle and Heavy Ion Transport code System (PHITS) code was used to simulate a cyclotron target and room. The dimensions of the room were 457 cm (length), 470 cm (width), and 320 cm (height). Gd₂O₃, B₄C, polyethylene (PE), and borated (5 wt% ^natB) PE with thicknesses of 5, 10, and 15 cm and their different combinations were selected as neutron absorbing materials. They were placed on the concrete walls to determine their effects on thermal neutrons. Thin B₄C and Gd₂O₃ were placed between the concrete wall and additional PE shield separately to decrease the required thickness of the additional shield, and the thermal neutron flux at certain depths inside the concrete was calculated for each condition. Subsequently, the optimum combination was determined with respect to radioactive waste reduction, price, and availability, and the total reduced radioactive concrete waste was estimated. Results and Discussion: In the specific conditions considered in this study, the front wall with respect to the proton beam contained radioactive waste with a depth of up to 64 cm without any additional shield. A single layer of additional shield was inefficient because a thick shield was required. Two-layer combinations comprising 0.1- or 0.4-cm-thick B₄C or Gd₂O₃ behind 10 cm-thick PE were studied to verify whether the appropriate thickness of the additional shield could be maintained. The number of transmitted thermal neutrons reduced to 30% in case of 0.1 cm-thick Gd₂O₃+10 cm-thick PE or 0.1 cm-thick B₄C+10 cm-thick PE. Thus, the thickness of the radioactive waste in the front wall was reduced from 64 to 48 cm. Conclusion: Based on price and availability, the combination of the 10 cm-thick PE+0.1 cmthick B₄C was reasonable and could effectively reduce the number of thermal neutrons. The amount of radioactive concrete waste was reduced by factor of two when considering whole concrete walls of the PET cyclotron vault.