• Journal of the Korea Convergence Society
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• v.9 no.7
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• pp.87-93
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• 2018

In this study, $^{99m}Tc$, $^{123}I$, $^{201}Tl$, $^{18}F$, and $^{131}I$, which are widely used in nuclear medicine, were transmitted through a bismuth shield. We investigated the shielding rates according to the type of radioisotope and the distance of measurement. For the experiment, 6 sheets of lead equivalent 0.25 mm Pb of bismuth shielding material were stacked one by one up to 1.50 mm as the thickness increased. The distance was 30 cm, 50 cm, and 100 cm, and the transmission dose was measured. As a result, the shielding rates was measured as the thickness increased, and the measured value decreased as the distance increased. The shielding rate of $^{123}I$ and $^{201}Tl$ was higher than $^{99m}Tc$, $^{18}F$ and $^{131}I$ showed lower shielding effect when there is a shielding material than when there is no shielding material due to high energy and ${\beta}$ rays. Based on the results of experiments, it would be helpful to reduce the exposure of nuclear medicine workers and to manage the exposure if bismuth shields are used depending on the type of radioisotope.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3854-3859
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• 2023

Background: This study aims to develop an evaluator that can quickly and accurately evaluate the shielding of low-energy industrial radiation generators. Methods: We used PyQt to develop a graphical user interface (GUI)-based program and employed the calculation methodology reported in the National Council on Radiation Protection and Measurements (NCRP)-49 for shielding calculations. We gathered the necessary factors for shielding evaluation using two libraries designed for Python, pandas and NumPy, and processed them into a database. We verified the effectiveness of the proposed program by comparing the results with those from safety reports of six domestic facilities. Results: After verifying the effectiveness of the program using the NCRP-49 example, we obtained an average error rate of 1.73%. When comparing the facility safety report and results obtained using the program, we found that the error rate was between 1.09% and 6.51%. However, facilities that did not use a defined shielding methodology were underestimated by 31.82% compared with the program (the final barrier thickness satisfied the shielding standard). Conclusion: The developed program provides a fast and accurate shielding evaluation that can assist personnel that work in radiation generator facilities and government officials in reviewing safety.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1210-1217
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• 2023

The assessment of radiation fields in the lithium loop pipes and dump tank during the operation were performed for International Fusion Materials Irradiation Facility - DEMO-Oriented NEutron Source (IFMIF-DONES) in order to obtain the radiation dose-rate maps in the component surroundings. Variance reduction techniques such as weight window mesh (produced with the ADVANTG code) were applied to bring the statistical uncertainty down to a reasonable level. The biological dose was given in the study, and potential shielding optimization is suggested and more thoroughly evaluated. The MCNP Monte Carlo was used to simulate a gamma particle transport for radiation shielding purposes for the current Li Systems' design. In addition, the shielding efficiency was identified for the Impurity Control System components and the dump tank. The analysis reported in this paper takes into account the radiation decay source from and activated corrosion products (ACPs), which is created by d-Li interaction. As a consequence, the radiation (resulting from ACPs and Be-7) shielding calculations have been carried out for safety considerations.

• Journal of Radiation Protection and Research
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• v.22 no.2
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• pp.77-83
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• 1997

Epoxy resin-type neutron shielding materials, KNS(Kaeri Neutron Shield)-101, KNS-102, and KNS-103 have been fabricated to be used in spent fuel shipping cask. The base material is epoxy resin, and polypropylene, aluminium hydroxide, and boron carbide are added. These shielding materials offer good fluidity at processing, which makes it possible to apply this resin shield to complicated geometric shapes such as shipping cask. The shielding property of these shielding materials for shipping cask for loading 28 PWR spent fuel assemblies has been evaluated. ANISN code is used to evaluate the shielding property of the shipping cask with the thickness of the three neutron shielding materials greater than 10 cm. As a result of analysis, the maximum calculated dose rate at the radial surface of the cask is determined to be $300{\mu}Sv/h$ and the maximum calculated dose rate at 100 cm from the cask is $97{\mu}Sv/h$. These dose rates remain within allowable values specified in related regulations.

• Journal of the Korean Society of Radiology
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• v.14 no.2
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• pp.149-155
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• 2020

Most of radiation protection clothing is made of lead with excellent radiation shielding because it has excellent process ability and economic efficiency and has a high atomic number. However, lead is classified as a hazardous heavy metal, and there is a risk of lead poisoning. Recently, research to replace lead has been actively conducted. In this study, a research on a shielding sheet with improved physical properties while maintaining the radiation shielding ability equivalent to that of conventional materials by mixing two materials that are harmless to the human body, such as BaSO₄ and Bi₂O₃, and a silicone material binder Was performed. For comparison evaluation with the existing lead shielding sheet, the shielding rate was evaluated using a 40 degree shielding sheet having the highest porosity. As a result, it was analyzed that the shielding rate was superior to 9 % or more at the same thickness. In addition, as a result of studies to improve the physical properties of the shielding sheet, it was analyzed that the shielding sheet mixed with BaSO₄/nylon/Bi₂O₃ was the best.

• The Korean Journal of Nuclear Medicine Technology
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• v.18 no.1
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• pp.110-114
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• 2014

Purpose: The advancement in PET/CT test devices has decreased the test time and popularized the test, and PET/CT tests have continuously increased. However, this increases the exposure dose of radiation workers, too. This study aims to measure the radiation shielding rate of $^{18}F-FDG$ with a strong energy and the shielding effect when worker wore an apron during the PET/CT test. Also, this study compared the shielding rate with $^{99m}TC$ to minimize the exposure dose of radiation workers. Materials and Methods: This study targeted 10 patients who visited in this hospital for the PET/CT test for 8 days from May 2nd to 10th 2013, and the $^{18}F-FDG$ distribution room, patient relaxing room (stand by room after $^{18}F-FDG$ injection) and PET/CT test room were chosen as measuring spots. Then, the changes in the dose rate were measured before and after the application of the APRON. For an accurate measurement, the distance from patients or sources was fixed at 1M. Also, the same method applied to $^{99m}TC's$ Source in order to compare the reduction in the dose by the Apron. Results: 1) When there was only L-block in the $^{18}F-FDG$ distribution room, the average dose rate was $0.32{\mu}Sv$, and in the case of L-blockK+ apron, it was $0.23{\mu}Sv$. The differences in the dose and dose rate between the two cases were respectively, $0.09{\mu}Sv$ and 26%. 2) When there was no apron in the relaxing room, the average dose rate was $33.1{\mu}Sv$, and when there was an apron, it was $22.3{\mu}Sv$. The differences in the dose and dose rate between them were respectively, $10.8{\mu}Sv$ and 33%. 3) When there was no APRON in the PET/CT room, the average dose rate was $6.9{\mu}Sv$, and there was an APRON, it was $5.5{\mu}Sv$. The differences in the dose and dose rate between them were respectively, $1.4{\mu}Sv$ and 25%. 4) When there was no apron, the average dose rate of $^{99m}TC$ was $23.7{\mu}Sv$, and when there was an apron, it was $5.5{\mu}Sv$. The differences in the dose and dose rate between them were respectively, $18.2{\mu}Sv$ and 77%. Conclusion: According to the result of the experiment, $^{99m}TC$ injected into patients showed an average shielding rate of 77%, and $^{18F}FDG$ showed a relatively low shielding rate of 27%. When comparing the sources only, $^{18F}FDG$ showed a shielding rate of 17%, and $^{99m}TC$'s was 77%. Though it had a lower shielding effect than $^{99m}TC$, $^{18}F-FDG$ also had a shielding effect on the apron. Therefore, it is considered that wearing an apron appropriate for high energy like $^{18}F-FDG$ would minimize the exposure dose of radiation workers.

• Journal of the Korea Institute of Building Construction
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• v.21 no.5
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• pp.459-468
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• 2021

Blocking electromagnetic waves on the exterior walls of buildings effectively reduces the intensity of electromagnetic fields in buildings, which leads to attenuation of electromagnetic disturbances, so there is a great interest in developing technologies. In this study, SiC by-products and graphite generated in the semiconductor field were selected and mixed into mortar after pretreatment such as pulverization to evaluate their physical properties and electromagnetic wave shielding performance. Considering the economic efficiency of each shielding material, only 10% volume of the outermost side of the experiment was mixed with each shielding material to evaluate the shielding performance. The shielding performance was predicted when the experiment was manufactured by mixing the shielding material with the entire volume of the experiment using the shielding effect evaluation formula. The results of the experiment showed that the shielding performance was up to 20 dB when SiC grains were mixed with shielding materials, the shielding performance was up to 18 dB when graphite powder was mixed with shielding materials, and the shielding performance was up to 28 dB when SiC powder was mixed with shielding materials, and the shielding performance was close to 30 dB, which is known to have a shielding rate of 99.9%.

• Journal of the Korean Society of Radiology
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• v.14 no.5
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• pp.545-552
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• 2020

The purpose of this study is to derive a lead thickness calculation formula for direct-shielded doors based on NCRP Report No.151 and IAEA Safety Report Series N0.47. After deriving the dose rate calculation formula for the direct shielded door, this formula was substituted for the lead shielding thickness calculation formula to derive the shielding thickness calculation formula at the door. The lead shielding thickness calculated from the derived direct shielded door shielding thickness calculation formula was about 6% lower than that calculated by the NCRP and IAEA secondary barrier shielding thickness calculation methods. This result is interpreted as meaning that the thickness calculation is more conservative from the NCRP and IAEA secondary barrier shielding thickness calculation methods and fits well for secondary beam shielding. In conclusion, it is thought that the formula for calculating lead shielding thickness of the direct shielded door derived in this study can be usefully used in the shield design of the door.

• The Journal of the Korea Contents Association
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• v.14 no.2
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• pp.457-466
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• 2014

During irradiation of lesions in cancer treatment with high energy electrons, normal tissue and critical organs are protected by the shielding material. Scattered radiation that generated the shielding materials affect the depth dose and atomic number. Therefore, we want to examine secondary particles and the scattered photons through calculation and its associated analysis, and compare the measurement for the aluminum, copper, and lead shielding substance of which thickness has 95% charge reduction. Dose change rate which effected scattering radiation was found to be +0.88% for material thickness, +0.43% for atomic number, and +19.70%, +15.20%, +12.40% for measurement, +25.00%, +15.10%, +13.70% for calculation on the aluminum, copper, and lead materials of which thickness has 95% charge reduction, respectively, As a result, we found that scattering rate was dependent on thickness than atomic number. In the dose increasing rate, scattered electrons are more important than scattered photon. For the above mentioned reasons, I think that high atomic number materials should be applied to reduce scattered radiation that generated with thickness effect.

• 전기의세계
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• v.29 no.1
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• pp.65-72
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• 1980

Arstrong-Whitehead theory on effective shielding and shielding failure is extended so that it is applicable to vertical line configuration the existing approaches on back-flash rate calculation are summerized and compared in particular on Model Line, and the usefulness of the whitehead approach is suggested.