• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.11a
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• pp.465-466
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• 2009

• Journal of the Korean Society of Radiology
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• v.11 no.5
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• pp.329-334
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• 2017

The purpose of this study is to estimating the possibility of manufacturing radiation shielding sheet by searching for environmentally friendly materials suitable for medical environment of medical radiation shielding. There are many tungsten products which are currently used as shielding materials in place of lead, but there are small problems in the mass production of lightweight shielding sheets due to economical efficiency. To solve these problems, a lightweight, environmentally friendly material with economical efficiency is required. In this study, Barium sulphate and Iodine were proposed. Both materials are already used as contrast medias in radiography, and it is predicted that the shielding effect will be sufficient in a certain region as a shielding material because of the characteristic of absorbing radiation. Therefore, in this study, we used a Monte Carlo simulation to simulate radiation shielding materials. When it is a contrast agent such as Barium sulfate and Iodine, the radiation absorption effect in the high energy region appears greatly, and the effectiveness of the two shielding substance in the energy region of the star with thickness of 120 kVp is also evaluated in the medical radiation imaging region. Simulated estimation results it was possible to estimate the effectiveness of shielding for all two substances. Iodine has higher shielding effect than barium sulfate, 0.05 mm thick appears great effect. Therefore, the Monte Carlo simulation confirms that iodine, which is a radiological contrast agent, is also usable as barium sulfate in the production of radiation shielding sheets.

• Journal of radiological science and technology
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• v.33 no.4
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• pp.363-367
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• 2010

In this paper we proposed an easy method to calculate the thickness of primary protective barrier for radiographic equipment. The concrete was selected for the shielding material. The area of protective barrier was divided into a controlled area and a noncontrolled area. For the computation of thickness, the data in NCRP Report 49 and 51 was used. For radiographic equipments whose maximum tubevoltages are 100 and 150 kVp, the thicknesses of concrete were calculated as a function of distance. From the calculated data, four analytical models were acquired by fitting an exponential decay function. From the equations acquired by this study, the thickness of primary protective barrier can be calculated approximately.

• The Proceeding of the Korean Institute of Electromagnetic Engineering and Science
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• v.2 no.2
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• pp.25-29
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• 1991

In this paper, laminated shielding effectiveness equation is derived from basic shielding theory and this equation is applied to calculate the shielding effectiveness for two typical non-magnetic shielding materials, Aluminium and steel, when they are coated with conductive paint.

• The Proceeding of the Korean Institute of Electromagnetic Engineering and Science
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• v.2 no.3
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• pp.17-25
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• 1991

Magnetic leakage flux which is generated from the levitation magnets, linear induction motors, and guide magnets of a MagLev(Magnetic Levitation) system is directly related to inter - system EMI, intra - system EMI, and biological effects. In this paper, the magnetic leakage flux from MagLev vechicles designed by Korea Resarch Institute of Ships & Ocean Engineering was calculated considering the various parameters which influence ma- gnetic field intensity around the MagLev system. Based on the calculated field intensity, the thickness of shielding material and shielding position for MagLev floor and side walls are calculated, taking into account the shielding effectiveness of a shield with minimum weight. For the nonuniform shielding method derived from the above procedure, the weight of a shield con- sisting of floor and side walls shielding can be reduced to more 50% than uniform shielding method.

• 전자공학회논문지 IE
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• v.44 no.2
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• pp.21-25
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• 2007

Multilayer transparent electromagnetic wave shielding film with 1 m wide, was fabricated by using roll to roll DC plasma coating with ITO and Ag layer on PET substrate. By optimizing properly the design parameters, such as a processing condition, the surface resistance and the thickness of each layers, the homogeneous film could be obtained. Electromagnetic wave shielding film showed the high shielding effectiveness of 23dB(99.5%) in 2-18 GHz range and the transmittance of 83.1% in 400-700nm.

• Journal of the Korean Society of Radiology
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• v.18 no.3
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• pp.283-292
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• 2024

Radiation side effects and complications on the ocular adnexa during electron beam therapy for orbital lymphoma can increase the incidence of posterior subcapsular cataracts. This study simulated a medical linear accelerator and a mathematical model of the eye using monte carlo simulations to evaluate the dose to the ocular adnexa and compare the shielding effectiveness on different parts of the ocular adnexa based on lens shield thickness. The dose assessment results of the ocular adnexa showed that the lens's sensitive area had the highest absorbed dose distribution when no shield was used, followed by the lens's non-sensitive area, the anterior chamber, vitreous humor, cornea, and eyelid in descending order. With the use of a shield, a 2 mm thick shield demonstrated a dose reduction effect of over 90% in the lens's sensitive area, over 83% in the non-sensitive area and anterior chamber, and a dose reduction effect of 30 to 62% in the vitreous body, cornea, and eyelid. For dose reduction in the lens's sensitive area during electron beam therapy for orbital lymphoma, it is necessary to use a shield of at least 2 mm thickness. Additionally, shielding strategies considering the thickness and area of the shield for other ocular adnexa besides the lens are required.

• 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.

• Journal of Radiation Protection and Research
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• v.26 no.2
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• pp.67-71
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• 2001

In order to shield the neutrons affecting the background of Low Level Gamma Ray Spectrometer, a neutron shielder was designed. The method used in this study for neutron shielding was the deceleration of fast neutrons by high density polyethylene(HDPE) and the absorption of those slowing-down neutrons by $B_4C$. The calculation results of neutron Interaction in HDPE using Monte Carlo simulation code MCNP4B showed that the thermal-neutron flux was maximum at 10 cm thickness of HDPE. The results also showed that 95% of the thermal neutrons were absorbed by 2 mm thickness of $B_4C$ absorber Consisted of 30 w% $B_4C$ and 70 w% polymer. The results of the Monte Carlo calculation were in good agreement with the experimental value obtained by a neutron shielding apparatus designed for this purpose.

• Journal of Radiation Protection and Research
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• v.36 no.1
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• pp.16-23
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• 2011

Accurate measurement of uranium enrichment is very important in nuclear material accountability. The analysis uncertainty of the uranium enrichment measurement with gamma-ray analysis was studied in the present work. FRAM (Fixed energy Response function Analysis with Multiple efficiencies) code was used to determine the uranium enrichment. If the shield materials were placed between the detector and the sample, the error was measured and analyzed. Measurement time was varied and the dependency of the analysis uncertainty on the measurement time was studied. Transmitted gamma-ray intensities and FWHMs of the peaks in the energy spectrum were measured as the shield thickness was varied. The transmitted gamma-ray intensity follows shape of the exponential function, and the FWHM was almost independent of the shield thickness. The uncertainty of FRAM analysis was studied when the thick shield material was placed between the detector and the sample. Our work could be helpful in analysis of the fissile material in uranium sample.