• Journal of Radiation Protection and Research
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• 제24권4호
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• pp.231-233
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• 1999

This report will introduce the epitome about the subject, HANARO (Hi-flux Advanced Neutron Application Reactor designed by KAERI in early 1995) Utilization Education Training Program Development and Operation, which is one of the nuclear research basic expansion businesses executed from 1999. 12. to strengthen the usage of HANARO. This program consists of the basic reactor experiments program for university students who have specialty of nuclear and other engineering, and the special research education program for faculties from universities and researchers from industrial fields. Principle lessons are reactor operation, radioisotope production, neutron activation analysis, neutron radiography, radiation shield (health physics), nuclear fuel combustion measurement by gamma scanning arrangement, and CNS (Compact Nuclear Simulator) and so on.

• 한국건설순환자원학회논문집
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• 제7권4호
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• pp.316-322
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• 2019

본 연구에서는 환경적·물리적 인자들의 불확실성을 반영하는 확률론적 접근법을 적용하여, 원자로 가동기간 동안 장시간 방사능에 노출된 원전 콘크리트 차폐벽의 재료적 특징 및 강도에 대한 영향을 평가하였다. 방사화에 따른 콘크리트의 재료적 특성 변화를 조사하였으며, 중성자 노출량과 시간과의 관계를 나타내는 중성자속 분석을 통해 차폐 콘크리트 의 시간의존적 압축강도와 인장강도의 변화를 예측하였다. 압축강도와 인장강도 각각의 변화에 따른 차폐 콘크리트의 파괴확률을 몬테카를로 시뮬레이션(Monte Carlo Simulation) 기법을 적용하여 추정하였다. 본 연구에서는 가동 40년 만인 2017년에 영구정지한 고리 1호기의 해체 안전성 평가를 위해, 이와 유사한 원전유형 및 관련 자료를 활용하여 콘크리트 생체차폐벽의 성능변화를 예측하였다.

• Nuclear Engineering and Technology
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• 제12권2호
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• pp.127-134
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• 1980

최근 가압경수로에서 압력용기 상단주변의 캐비티에 중성자가 새어나오는 사실이 판명되자 이에 대한 차폐문제가 심각하게 논의되기 시작했다. 본 논문에서는 현재 운전중인 원자로에서 이것을 어떻게 해결하고 있는가를 예시하였다. 예를들면 붕소가 들어있는 주머니를 쌓아 올리는 것에서부터 화폐구조물을 영구히 설치하는 것에 이르기까지 여러가지 방법으로 중성자흐름을 막는 대책을 논의하였다. 결론적으로 이 문제해결을 위한 가장 현실적이고 가능한 차폐설계안 몇가지를 제시하였다. 특히 그중에서도 중성자가 어떤 경로로 흘러 나오는가의 규명. 차폐자료의 선정, 차폐설계의 특성과 외형문제 그리고 각 설계안을 검토키 위한 수학적 모델 제시에 역점을 두었다.

• Journal of Radiation Protection and Research
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• 제26권2호
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• pp.67-71
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• 2001

극저준위 방사능측정시스템의 백그라운드에 영향을 주는 중성자를 차폐하기 위한 차폐체를 설계하였다. 중성자 차폐방법은 고 밀도 폴리에틸렌을 이용하여 고속중성자를 감속한 후 $B_4C$를 이용하여 감속된 열중성자를 흡수하는 방법을 이용하였다. 몬테카를로 모사방법인 MCNP4B 코드를 이용하여 계산한 결과 고 밀도 폴리에틸렌의 두께가 10 cm 일 때 열중성자속이 최대가 되는 것으로 나타났으며 감속된 중성자의 흡수는 용제에 자연상태의 $B_4C$ 분말을 30 w% 섞을 경우 2 mm의 두께에서 94%의 중성자 흡수가 일어나는 것으로 나타났다. 또한 몬테카를로 모사를 통한 계산결과의 타당성 여부를 조사하기 위하여 중성자 차폐실험 장치를 제작하여 실험 결과와 비교하였으며, 비교 결과 실험값과 일치하는 것으로 나타났다.

• Journal of Radiation Protection and Research
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• 제42권3호
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• pp.141-145
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• 2017

Background: The concrete walls inside the vaults of cyclotron facilities are activated by neutrons emitted by the targets during radioisotope production. Reducing the amount of radioactive waste created in such facilities is very important in case they are decommissioned. Thus, we proposed a strategy of reducing the neutron activation of the concrete walls in cyclotrons during operation. Materials and Methods: A polyethylene plate and B-doped Al sheet (30 wt% of B and 2.5 mm in thickness) were placed in front of the wall in the cyclotron room of a radioisotope production facility for pharmaceutical use. The target was Xe gas, and a Cu block was utilized for proton dumping. The irradiation time, proton energy, and beam current were 8 hours, 30 MeV, and $125{\mu}A$, respectively. To determine a suitable thickness for the polyethylene plate set in front of the B-doped Al sheet, the neutron-reducing effects achieved by inserting such sheets at several depths within polyethylene plate stacks were evaluated. The neutron fluence was monitored using an activation detector and 20-g on de Au foil samples with and without 0.5-mm-thick Cd foil. Each Au foil sample was pasted onto the center of a polyethylene plate and B-doped Al sheet, and the absolute activity of one Au foil sample was measured as a standard using a Ge detector. The resulting relative activities were obtained by calculating the ratio of the photostimulated luminescence of each foil sample to that of the standard Au foil. Results and Discussion: When the combination of a 4-cm-thick polyethylene plate and B-doped Al sheet was employed, the thermal neutron rate was reduced by 78%. Conclusion: The combination of a 4-cm-thick polyethylene plate and B-doped Al sheet effectively reduced the neutron activation of the investigated concrete wall.

• 한국건축시공학회지
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• 제17권1호
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• pp.9-14
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• 2017

본 연구에서 비정질 붕소강 섬유를 혼입한 콘크리트의 역학적 성능 및 중성자 차폐성능을 평가하였다. 비정질 붕소강 섬유를 콘크리트 체적 대비 0.25%에서 1.0%까지 혼입하여 굳지 않은 콘크리트의 공기량과 슬럼프값, 경화된 콘크리트의 압축강도, 휨강도, 휨인성 및 중성자 차폐성능을 평가하였다. 실험결과, 비정질 붕소강 섬유의 혼입량이 증가할수록 콘크리트의 휨인성 및 중성자 차폐성능이 향상되는 것으로 나타났다. 이를 통해 비정질 붕소강 섬유의 혼입이 중성자 차폐성능 뿐만 아니라 역학적 성능을 효과적으로 개선시켜 줄 것이라고 기대된다.

• 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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
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• pp.618-623
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• 1998

A shielding analysis was performed for the end shield of CANDU 6 reactor. The one-dimensional discrete ordinate code ANISN with a 38-group neutron-gamma library, extracted from DLC-37D library, was used to estimate the dose rate for the natural uranium CANDU reactor. For comparison MCNP-4B calculation was performed for the same system using continuous, discrete and multi-group libraries. The comparison has shown that the total dose rate of the ANISN calculation agrees well with that of the MCNP calculation. However, the individual dose rate (neutron and gamma) has shown opposite trends between AMISN and MCNP estimates, which may require a consistent library generation for both codes.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3816-3823
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• 2021

The mechanical features evaluated theoretically using Makishima-Mackenzie's model for glasses xBaO-(50-x) PbO-50P₂O₅ where x = 0, 5, 10, 15, 20, 30, 40, and 50 mol%. Wherefore, the elastic characteristics; Young's, bulk, shear, and longitudinal modulus calculated. The obtained result showed an increase in the calculated values of elastic moduli with the replacement of the PbO by BaO contents. Moreover, the Poisson ratio, micro-hardness, and the softening temperature calculated for the investigated glasses. Besides, gamma and neutron shielding ability evaluated for the barium doped lead phosphate glasses. Monte Caro code (MCNP-5) and the Phy-X/PSD program applied to estimate the mass attenuation coefficient of the studied glasses. The decrease in the PbO ratio has a negative effect on the MAC. The highest MAC decreased from 65.896 cm²/g to 32.711 cm²/g at 0.015 MeV for BPP0 and BPP7, respectively. The calculated values of EBF and EABF showed that replacement of PbO with BaO contents in the studied BPP glasses helps to reduce the number of photons accumulated inside the studied BPP glasses.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2614-2624
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• 2022

Monte Carlo simulations were used to model a portable Neutron backscattering (NBT) sensor suitable for detecting plastic anti-personnel mines (APMs) buried in dry and moist soils. The model consists of a 100 MBq ²⁵²Cf source encapsulated in a neutron reflector/shield assembly and centered between two ³He detectors. Multi-parameter optimization was performed to investigate the efficiency of Be/Zr(BH₄)₄ and Be/Be(BH₄)₂ assemblies in terms of increasing the signal-to-background (S/B) ratio and reducing the total dose equivalent rate. The MCNP results showed that 2 cm Be/3 cm Zr(BH₄)₄ and 2 cm Be/3 cm Be(BH₄)₂ are the optimal configurations. However, due to portability requirements and abundance of Be, the ²⁵²Cf-2 cm Be/3 cm Be(BH₄)₂ NBT model was selected to scan the center of APM buried 3 cm deep in dry and moist soils. The selected NBT model has positively identified the APM with a S/B ratio of 886 for dry soils of 1 wt% hydrogen content and with S/B ratios of 615, 398, 86, and 12 for the moist soils containing 4, 6, 10, and 14 wt% hydrogen, respectively. The total dose equivalent rate reached 0.0031 mSv/h, suggesting a work load of 8 h/day for 806 days within the permissible annual dose limit of 20 mSv.