• Journal of radiological science and technology
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• v.46 no.3
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• pp.187-195
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• 2023

A shield was made by mixing materials such as bismuth(Bi) and barium(Ba) with silicon to evaluate its shielding ability. Bismuth was made into a shield by mixing a bismuth oxide(Bi₂O₃) colloidal solution and a silicon base and applied to a fibrous fabric, and barium was made by mixing lead oxide(PbO) and barium sulfate(BaSO₄) with a silicon curing agent and solidifying it to make a shield. The test was conducted according to the lead equivalent test method for X-ray protective products of the Korean Industrial Standard. The experiment was conducted by increasing the shielding body one by one from the test condition of 60 kVp, 200 mA, 0.1sec and 100 kVp, 200 mA, 0.1 sec. At 60 kVp, 2 lead oxide-barium sulfate shields, 2 bismuth oxide 1.5 mm shields, and 5 bismuth oxide 0.3 mm shields showed shielding ability equal to or higher than that of lead 0.5 mm. At 100 kVp, 2 lead oxide-barium sulfate shields and 2 bismuth oxide 1.5 mm shields showed shielding ability equal to or higher than that of lead 0.5 mm. It was confirmed that when using 2 pieces of lead oxide-barium sulfate and 1.5 mm of bismuth oxide, respectively, it has shielding ability equivalent to that of lead. Bismuth oxide and lead oxide-barium sulfate are lightweight and have excellent shielding ability, thus they have excellent properties to be used as an apron for radiation protection or other shielding materials.

• Journal of the Korean Society of Radiology
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• v.13 no.7
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• pp.969-977
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• 2019

Applying the bismuth shield used to reduce the radiation exposure, image quality may be reduced due to beam hardening caused by the shield during CT scan. Therefore, we tried to find out the energy range that can reduce image degradation by applying GSI mode of G company's dual energy CT and examine the possibility through experiment. As a result, after bismuth shielding, 118 ± 10.6 HU and 50.1 ± 14.6 HU at 50 keV after dual-energy CT scan were the most similar to the CT value before image deterioration(p> 0.05). It was measured 176.6 ± 7.1 and 138.3 ± 1.1 at 50 keV(p> 0.05). Experiments showed that the use of the shield during CT inspection inevitably degrades the image quality, but experiments show that the GSI function of the dual energy CT can maintain the image quality even when the shield is used. If the various shields are secured after the evaluation using the dual energy CT, it is expected to overcome the disadvantages of poor image quality caused by the use of the radiation shield for reducing the exposure, which is the biggest disadvantage of the CT scan.

• Journal of the Korean Society of Radiology
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• v.16 no.6
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• pp.687-695
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• 2022

To find a 3D printer material that can replace lead used as a shield for high-energy electron beam treatment, the shielding composites were simulated by using MCNP6 programs. The Percent Depth Dose (PDD), Flatness, and Symmetry of linear accelerators emitting high-energy electron beams were measured, and the linear accelerator was compared with MCNP6 after simulation, confirming that the source term between the actual measurement and simulation was consistent. By simulating the lead shield, the appropriate thickness of the lead shield capable of shielding 95% or more of the absorbed dose was selected. Based on the absorption dose data for lead shield with a thickness of 3 mm, the shielding performance was analyzed by simulating 1, 5, 10, and 15 mm thicknesses of ABS+W (10%), ABS+Bi (10%), and PLA+Fe (10%). Each prototype was manufactured with a 3D printer, measured and analyzed under the same conditions as in the simulation, and found that when ABS+W (10%) material was formed to have a thickness of at least 10mm, it had a shielding performance that could replace lead with a thickness of 3mm. The surface morphology and atomic composition of the ABS+W (10%) material were evaluated using a scanning electron microscope (SEM) and an energy dispersive X-ray spectrometer (EDS). From these results, it was confirmed that replacing the commercialized lead shield with ABS+W (10%) material not only produces a shielding effect such as lead, but also can be customized to patients using a 3D printer, which can be very useful for high-energy electron beam treatment.

• Journal of the Korean Society of Radiology
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• v.13 no.1
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• pp.87-93
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• 2019

Recently, the number of passengers using airplanes is rapidly increasing due to the increase of overseas travelers. Therefore, the probability of exposure to natural radiation due to altitude is increasing due to the increase in flight time. Cosmic-ray penetrates the Earth's magnetic field belt Van Allen, which is located at an altitude of 400 km to 1200 km. Most cosmic rays are blocked at Van Allen belt. However, cosmic-ray could be not completely blocked, and a small amount of cosmic-ray affects the earth. In general, if the altitude was increased by 100m, the natural exposure dose increased by 0.03 mSv on the Earth. In this study, I tried to minimize the exposure to natural radiation in airplanes when boarding airplanes. Especially, I was aimed to minimize radiation exposure by protecting the highly sensitive thyroid gland among human organs. According to the results of the study, the designed shielding handkerchief was able to shield cosmic natural radiation dose by more than 70%. In conclusion, the application of the shielding handkerchief made in this study can be effectively shield natural radiation.

• Proceedings of the Korean Vacuum Society Conference
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• 2012.02a
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• pp.582-582
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• 2012

Characteristics of a fusion-driven transmutation reactor was investigated. A compact reactor concept is desirable from an economic viewpoint. For the optimal design of a reactor, a radial build of reactor components has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor components. In a transmutation reactor, design of blanket and shield play a key role in determining the size of a reactor; the blanket should produce enough tritium for tritium self-sufficiency, the transmutation rate of waste has to be maximized, and the shield should provide sufficient protection for the superconducting toroidal field (TF) coil. To determine the radial build of the blanket and the shield, not only a radiation transport analysis but also a burnup calculation were coupled with the system analysis and it allowed the self-consistent determination of the design parameters of a transmutation reactor.

• 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 radiological science and technology
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• v.47 no.3
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• pp.205-211
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• 2024

Currently, with the development of technologies, X-ray examinations for medical examinations at hospital is increasing. This study was conducted to help reduce radiation exposure by measuring the exposure dose received by pediatric patients and the spatial dose of the X-ray room. Dosimeters were installed in the eyeball, thyroid gland, breast, gonads and 4 directions at a distance of 30 cm, 40 cm, 50 cm from the phantom. The dose was measured ten times each, before, and after the application of the bismuth shield under the examination conditions of the head, chest, and abdomen of pediatric patients. Under the condition of head examination, when a shielding was applied, the dose reduction rate was 68.58% for the eyeball, 72.88% for the thyroid, 84.2% for the breast, and 72.36% for the gonad. The chest examination showed reductions of 19.56% eyeball, 56.98% thyroid, 1.21% breast, and 0.68% gonad. The abdominal examination showed reduction rates of 2.6% eyeball, 10.67% thyroid, 19.85% breast, and 82.02% gonad. Spatial dose decreased by 62.25% at 30 cm, 61.16% at 40 cm, and 68.68% at 50 cm. When the bismuth shield was applied, there was a decrease in dose across all examinations, as well as a reduction in spatial dose. Continued research on the use of bismuth shields will help radiological technologists achieve their goal of dose reduction.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.25 no.6
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• pp.959-964
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• 2001

An experimental investigation on the near-field electromagnetic loss of thin copper layers has been presented using microfabricated microwave transceivers for applications to multi-chip microsystems. Copper layers in the thickness range of 0.2$\mu$m∼200$\mu$m have been electroplated on the Pyrex glass substrates. Microwave transceivers have been fabricated using the 3.5mm$\times$3.5mm nickel microloop antennas, electroformed on the silicon substrates. Electromagnetic radiation loss of the copper layers placed between the microloop transceivers has been measured as 10dB∼40dB for the wave frequency range of 100MHz∼1GHz. The 0.2$\mu$m-thick copper layer provides a shield loss of 20dB at the frequencies higher than 300MHz, whereas showing a predominant decreases of shield loss to 10dB at lower frequencies. No substantial increase of the shield effectiveness has been found for the copper shield layers thicker that 2 $\mu$m.

• Journal of Radiation Protection and Research
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• v.45 no.3
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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.

• Journal of the Korean Geotechnical Society
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• v.39 no.10
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• pp.49-55
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• 2023

The discovery of lunar ice in the lunar polar region has fueled international interest in in situ resource utilization (ISRU) and the construction of lunar habitats. Unlike Earth's atmosphere, the Moon presents unique challenges, including frequent meteoroid impacts, direct exposure to space radiation, and extreme temperature variations. To safeguard lunar habitats from these threats, the construction of a protective shield is essential. Lunar regolith, as a construction material, offers distinct advantages, reducing transportation costs and ensuring a sustainable supply of raw materials. Moreover, it streamlines manufacturing, integration schedules, and enables easy repairs and modifications without Earth resupply. Adjusting the shield's thickness within the habitat's structural limits remains feasible as lunar conditions evolve. Although extensive research on protective shields using lunar regolith has been conducted, unresolved conflicts persist regarding shield requirements. This study conducts a comprehensive analysis of the primary lunar threats and suggests a minimum shield thickness of 2 m using lunar regolith. Furthermore, it outlines the necessary technology for the rapid construction of such protective shields.