• Nuclear Engineering and Technology
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• 제56권7호
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• pp.2767-2774
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• 2024

In this study, investment casting shell waste successfully utilized to produce cordierite/mullite/zircon composites. Green pellets, consisting of investment casting shell waste, alumina, and magnesia, were prepared and sintered at temperatures between 1250 and 1350 ℃. The influence of the sintering temperature on the crystalline phase composition, densification behavior, flexural strength, microstructure, and radiation shielding properties of the cordierite/mullite/zircon composites is investigated. Phase analysis showed that characteristic cordierite peaks appear at 1250 ℃, but the complete conversation of silica from investment casting shell waste into cordierite requires a sintering temperature of at least 1300 ℃. Notably, the cordierite/mullite/zircon composite sintered at 1350 ℃ exhibited a sixfold increase in flexural strength compared to the ceramic composite directly fabricated from investment casting shell waste at the same sintering temperature. Furthermore, the effect of Y₂O₃ addition on composites' radiation shielding properties is investigated. The results show that the Y₂O₃ addition improves densification behavior, enhancing the shielding capabilities of the composites against fast neutron and gamma radiation. Our findings suggest that the developed ceramic composites show significant potential for gamma-ray and neutron shielding applications.

• Nuclear Engineering and Technology
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• 제56권7호
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• pp.2444-2451
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• 2024

This work demonstrates a new sol-gel approach for synthesizing PbO₂-doped zirconia using zircon mineral precursors. The streamlined methodology enables straightforward fabrication of the doped zirconia composites. Comprehensive materials characterization was performed using XRD, SEM, and TEM techniques to analyze the crystal structure, microstructure, and morphology. Quantitative analysis of the XRD data provided insights into the nanoscale crystallite sizes achieved, along with their relationship to lattice imperfections. Furthermore, the gamma-ray shielding capacity for the PbO₂-doped zirconia samples was estimated by the Monte Carlo simulation, which proves an increase in the gamma ray shielding properties by raising the Pb concentration. The linear attenuation coefficient increased between 0.467 and 0.499 cm^-1 (at 0.662 MeV) by increasing the Pb content between 11 and 21 wt%. By increasing the Pb content to 21 wt%, the synthesized composites' lead equivalent thickness reaches 2.49 cm. The radiation shielding properties for the synthesized composites revealed a remarkable performance against low and intermediate γ-ray photons, with radiation shielding capacity of 37.3 % and 21.4 % at 0.662 MeV and 2.506 MeV, respectively. As a result, the developed composites can be employed as an alternative shielding material in hospitals and radioactive zones.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.278-284
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• 2023

The synthetic B₂O₃-As₂O₃ glass ceramic are prepared to investigate the physical properties and the radiation shielding capabilities with the variation of concentration of the As₂O₃ with 10, 20, 30, and 40%, respectively. XRD analyses are performed on the fabricated glass-ceramic and depicted the improvement of crystallinity by adding As₂O₃. The radiation shielding properties are studied for the B₂O₃-As₂O₃ glass ceramic. The values of linear attenuation coefficient (LAC) are varied with the variation of incident photon gamma energy (23.1-103 keV). The LAC values enhanced from 12.19 cm^-1-37.75 cm^-1 by raising the As₂O₃ concentration from 10 to 40 mol% at low gamma energy (23.1 keV) for BAs10 and BAs40, respectively. Among the shielding parameters, the half-value layer, transmission factor, and radiation protection efficiency are estimated. Furthermore, the fabricated samples of glass ceramic have low manufacturing costs and good shielding features compared to the previous work. It can be concluded the B₂O₃-As₂O₃ glass ceramic is appropriate to apply in X-ray or low-energy gamma-ray shielding applications.

• 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.

• Journal of Radiation Protection and Research
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• 제22권2호
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• pp.77-83
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• 1997

사용 후 핵연료 수송용기 등에 사용되는 에폭시수지계 중성자 차폐재, KNS(Kaeri Neutron Shield)-101, KNS-102 및 KNS-103를 제조하였다. 기본물질은 에폭시수지이며, 첨가제로는 폴리프로필렌, 수산화알루미늄 및 탄화붕소이다. 이들 중성자 차폐재들은 유동성이 좋아 수송용기와 같은 복잡한 구조에 사용할 수 있다. 제조된 중성자 차폐재들을 가압경수로 사용 후 핵연료 28다발을 수송할 수 있는 수송용기에 적용하여 차폐능 평가를 수행하였다. 세가지 중성자 차폐재를 수송용기에 적용하여 ANISN 코드로 차폐능 평가를 수행한 결과 정상수송시 중성자 차폐재의 두께가 10 cm 이상 일때 수송용기 반경방향표면에서 최대 방사선량율은 $300{\mu}Sv/h$로 나타났으며, 수송용기 표면에서 100 cm 지점에서의 최대 방사선량율은 $97{\mu}Sv/h$로 나타났다. 이들은 모두 관련된 법규들에서 규정된 최대허용 방사선량율을 만족하는 것으로 나타났다.

• Journal of Radiation Protection and Research
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• 제41권3호
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• pp.310-314
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• 2016

Background: FCTC10 container is designed to transport $^{60}Co$ radioactive sources used in irradiation industry. It belongs to Type B(U) Category III (yellow) package when being loaded with a $^{60}Co$ source of $1.8{\times}10^5$ Ci. Materials and Methods: The container is constituted of shielding container, basket, protective cover and bracket. Shielding ability is provided mainly by stainless steel shells, tungsten alloy and lead among steel shells. Radiation level around the container has been calculated with both Monte Carlo simulations and measurements. Results and Discussion: It is proven that the shielding performance of the container fulfills the requirements in GB11806-2004 (Regulations for the safe transport of radioactive material, China Standard Press). Exposure doses to workers and to critical groups of public were calculated based on hypothetical exposure scene according to transport practice experience. Conclusion: The results show that doses to workers and public are less than the constraint dose considered in design, and the radiation level would be increased less than a factor of 2 under design basis accidents.

• Nuclear Engineering and Technology
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• 제56권8호
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• pp.3378-3384
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• 2024

Investigations were conducted on the addition of barium's impact on the radiation shielding and physical attributes of five different glasses, designated S1-S5, with varying BaO contents. Using two point sources namely Co⁶⁰ and Cs¹³⁷ along with a scintillation detector [NaI(TL)], experimental measurements were made of the shielding parameters of γ-rays, namely the effective atomic number (Z_eff), electron density (N_el), half-value layer (HVL), linear attenuation coefficient (μ), mass attenuation coefficient (μ_m), mean free path (λ), and radiation protection effectiveness at the energies of 0.664, 1.177, and 1.334 MeV, and comparisons made with recently considered glasses as well as frequently employed materials for γ-ray shielding. The results show that the examined glasses' physical and radiation shielding qualities are improved by the addition of BaO. The μ values increased from 0.245 to 0.275 cm^-1 (0.662 MeV), from 0.174 to 0.198 cm^-1 (1.173 MeV), and from 0.161 to 0.189 (1.332 MeV). The observed values of HVL decreased from 2.83, 3.98, and 4.3 cm to 2.5, 3.5, and 3.62 cm at 0.662, 1.173, and 1.332 MeV, respectively, for the samples S1 and S5. In addition, the S5 glass sample was determined to have the best protection against photon among all the samples that were evaluated, as well as against recently considered glasses and those materials often utilized for gamma-ray shielding purposes.

• 대한방사선기술학회지:방사선기술과학
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• 제7권1호
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• pp.35-40
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• 1984

Side-direction scattered dose from various radiation shielding materials was measured at 50cm distance from the central beam of primary ray by used several kinds of added filters for a x-ray deep therapeutic installation, the obtained results were as follows : 1. Dose rate by tube voltage was more increased at heavy filtration than light filtration. 2. Scattered doses produced by constant tube voltage in all shielding materials were decreased at heavier filtration. 3. Scattered doses produced by constant shielding material in all tube voltages were decreased at heavier filtration.

• 한국방사선학회논문지
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• 제17권3호
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• pp.441-447
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• 2023

현재 의료방사선 분야에서 차폐를 목적으로 주로 사용되고 있는 납(Pb)소재는 방사선 차폐 기능은 뛰어 나지만 납 자체가 가지고 있는 인체 유해성과 무거운 무게에 의한 불편함 때문에 지속적으로 직, 간접적으로 방사선 피폭 위험을 차단함과 동시에 납 소재를 대체할 수 있는 인체 친화적인이며 가벼우면서 사용편의성을 가진 차폐소재의 연구는 지속적으로 진행되어지고 있다. 본 연구에서는 일반적으로 사용되는 PET(polyethylene terephthalate) 필름과 실제 방사선 방호복 사용되는 원단소재를 기재로 하여 방사선을 차폐할 수 있는 금속물질인 비스무트, 텅스텐, 주석을 스퍼터링 진공증착 방식을 통한 다층박막을 구현하여 차페필름을 제작하여 방사선 차폐소재로의 적용가능성을 평가하였다. 차폐필름을 제작하기 위한 인가전압, 롤 구동속력, 가스공급량을 제어하면서 차폐물질별 최적화된 조건을 확립하여 방사선 차폐필름 제작하였다. 모재와 차폐금속박막간 밀착력 확인은 Cross-cut 100/100으로 확인하였고 시간에 따른 박막의 변화를 측정하기 위해 내열탕 테스트 1시간을 통하여 박막의 안정성을 확인하였다. 최종적으로 구현된 차폐필름의 차폐성능은 한국방사선진흥협회를 통한 실제 방사선 차폐성능을 측정한 결과 시험조건(역넓은 빔, 관전압 50 kV, 반가층 1.828 mmAl)을 설정하여 감쇠비 16.4 (초기값 0.300 mGy/s, 측정값 0.018 mGy/s)와 감쇠비 4.31(초기값 0.300 mGy/s, 측정값 0.069 mGy/s)의 결과를 얻었다. 추후 제품화를 위한 공정효율성을 확보하여 가벼우면서 차폐성능을 보유한 필름 및 원단을 활용하여 방사선 방호복이나 차폐기능을 가진 건축자재로의 필름적용을 위한 초석을 마련하였다.

• 한국융합학회논문지
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• 제9권12호
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• pp.99-106
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• 2018

의료방사선 방어를 위해 사용되는 차폐 시트의 제조과정에서 인장강도를 유지하면서 차폐 재료의 충전율을 높여 차폐 성능을 향상시키는 방안을 제시하고자 한다. 본 연구에서 제안된 개질제는 고분자 수지재료와의 융합에 있어서 분자량을 높여 재료의 친화성을 높이는 역할을 수행한다. 차폐시트의 산화텅스텐의 충전율과 인장강도, 차폐 성능 등을 평가하였다. 공정과정에서 개질제는 분자량과 밀도를 증가시켰고, 성형 과정에서 퍼짐 형상이 적용되었고, 성능 향상과 재료와의 친화성, 인장강도를 유지하기 위해 사용된 개질제 PMMA는 20%를 혼합할 경우 가장 우수한 효과를 얻을 수 있다. 본 연구에서 제시된 재료의 융합과 개질제를 통해 차폐시트의 대량생산이 가능하며, 향후 경량의 차폐복 제작에 기여할 것입니다.