• Journal of Radiation Protection and Research
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• 제43권1호
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• pp.20-28
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• 2018

Background: Heat energy produced in nuclear reactors and nuclear fuel cycle facilities interactions modifies the physical properties of the shielding materials containing water content. Therefore, in the present paper, effect of the heat on shielding effectiveness of the concretes is investigated for gamma and neutron. The mass attenuation coefficients, effective atomic numbers, fast neutron removal cross-section and exposure buildup factors. Materials and Methods: The mass attenuation coefficients, effective atomic numbers, fast neutron removal cross-section and exposure buildup factors of ordinary and heavy concretes were investigated using NIST data of XCOM program and Geometric Progression method. Results and Discussion: The improvement in shielding effectiveness for photon and reduction in fast neutron for ordinary concrete was observed. The change in the neutron shielding effectiveness was insignificant. Conclusion: The present investigation on interaction of gamma and neutron radiation would be very useful for assessment of shielding efficiency of the concrete used in high temperature applications such as reactors.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2017년도 추계 학술논문 발표대회
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• pp.155-156
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• 2017

Borosilicate glass was incorporated to improve the neutron shielding capability of concrete. Boron is a typical neutron shielding material, and it is contained in borosilicate glass. However, borosilicate glass causes alkali-silica reaction, which damages the concrete. Therefore, studied to reduce the expansion due to alkali-silica reaction and to improve the neuton shielding capability. The measurement of the expansion due to the alkali-silica reaction was based on ASTM C 1260. Experimental results show that the expansion due to alkali-silica reaction is reduced when borosilicate glass powder incorporated. In addition, the neutron shielding capability was significantly improved when the fine aggregate replaced with borosilicate glass.

• 한국건설순환자원학회논문집
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• 제4권3호
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• pp.269-275
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• 2016

본 연구에서는 붕규산 유리와 비정질 붕소강 섬유를 혼입한 콘크리트의 역학적 성능 및 중성자 차폐성능을 평가하였다. 잔골재를 붕규산 유리로 치환한 콘크리트와 비정질 붕소강 섬유를 보강한 콘크리트를 제조하여 압축강도, 정탄성계수, 압축인성, 휨강도, 휨인성, 중성자 차폐성능을 평가하였다. 실험결과, Plain 콘크리트와 대비하여 붕규산 유리를 혼입한 콘크리트의 경우 압축강도 및 휨강도는 저하되었으나, 중성자 차폐성능은 향상되는 것으로 나타났다. 비정질 붕소강 섬유를 혼입한 콘크리트의 경우 Plain 콘크리트와 대비하여 압축인성 및 휨인성이 증진되었으며 중성자 차폐성능도 향상되는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.647-653
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• 2020

Stainless steel is used commonly in nuclear applications for shielding radiation, so in this study, three different types of new stainless steel samples were designed and developed. New stainless steel compound ratios were determined by using Monte Carlo Simulation program Geant 4 code. In the sample production, iron (Fe), nickel (Ni), chromium (Cr), silicium (Si), sulphur (S), carbon (C), molybdenum (Mo), manganese (Mn), wolfram (W), rhenium (Re), titanium (Ti) and vanadium (V), powder materials were used with powder metallurgy method. Total macroscopic cross sections, mean free path and transmission number were calculated for the fast neutron radiation shielding by using (Geant 4) code. In addition to neutron shielding, the gamma absorption parameters such as mass attenuation coefficients (MACs) and half value layer (HVL) were calculated using Win-XCOM software. Sulfuric acid abrasion and compressive strength tests were carried out and all samples showed good resistance to acid wear and pressure force. The neutron equivalent dose was measured using an average 4.5 MeV energy fast neutron source. Results were compared to 316LN type stainless steel, which commonly used in shielding radiation. New stainless steel samples were found to absorb neutron better than 316LN stainless steel at both low and high temperatures.

• 한국재료학회지
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• 제9권4호
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• pp.404-408
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• 1999

Effects of heating time and radiation under high temperature on the properties of thermal, mechanical and shielding ability of modified (KNS-101), hydrogenated bisphenol-A(KNS-201) type epoxy resin and phenol-novolac(KNS-301) type epoxy resin based neutron shielding materials that are used for shipping casks for radioactive material have been investigated. At early stages, the offset temperatures of KNS-101, KNS-201 and KNS-301 increased with the heating time under high temperature, but it was rarely affected by the heating time in the later stages. In addition, the thermal conductivities of KNS-101 and KNS-201 decreased with heating time, but that of KNS-301 increased with the heating time. On the contrary, the thermal expansion coefficients of neutron shielding materials decreased with heating time. At the high temperature, the tensile strength and flexural strength of the shielding materials decreased with heating time. On the contrary, the thermal expansion coefficients of neutron shielding materials decreased with heating time. At the high temperature, the tensile strength and flexural strength of the shielding materials of KNS-101 and KNS-301 increased with heating time, but those of KNS-201 decreased with heating time. The shielding ability of neutron shielding materials slightly increased with the radiation dose, and shielding abilities of shielding materials of KNS-101 and KNS-201 were affected to a more extent than that of KNS-301 by radiation dose under high temperature.

• 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.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2018년도 추계 학술논문 발표대회
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• pp.114-115
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• 2018

To develop an effective shielding material for spent fuel that emits fast neutrons is necessary. In this study, thermal neutron and fast neutron shielding performance of polyethylene coated boron carbide-incorporated cement paste was quantitatively analyzed by Monte Carlo N-Particle transport code (MCNP) simulations. As the results of the simulations, fast neutrons were effectively shielded through large quantity of hydrogen and boron elements in polyethylene and boron carbide.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3448-3451
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• 2022

Biochar is a novel carbon based material derived from waste that shows promising properties for several applications. In this paper we investigate its potential use as a low cost, greener alternative to commonly used aggregates employed to enhance the neutron shielding performance of concrete. Monte Carlo simulations are performed with the PHITS code to estimate the neutron attenuation of blank and biochar-reinforced concrete exposed to high energy neutrons. We find that the shielding performance of concrete with 15% biochar is comparable with commonly used materials such as Boron Carbide at 20% and exceeds that of Basalt fibers with the same concentration, making these composites an interesting greener alternative to current solutions. A combination of biochar and heavier fillers also show extremely promising performance.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1541-1558
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• 2023

The new Modified-690Gd alloy, namely as Ni-30Cr-(10-x) Fe-xGd (x = 0.5, 1.0, 1.5,2.0, 3.0 wt%) for function structure integrated thermal neutron shielding has been prepared and characterized. The Modified-690Gd alloy was mainly composed of γ austenite matrix and (Ni, Cr, Fe)₅Gd precipitated along grain boundaries. The new Modified-690Gd alloy had great mechanical properties, which had the tensile strength exceeding 620 MPa and the elongation being above 50%. Meanwhile, this alloy had excellent weldability and good corrosion resistance in boric acid. The new Modified-690Gd alloy is expected to be a kind of high efficiency thermal neutron shielding materials.