• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.847-852
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• 1995

A shielding analysis for KSC-7, the shipping cask for transporting the 7 PWR spent fuel assemblies, has been carried out. Radiation source term has been calculated on spent fuel with burnup of 50,000 MWD/MTU and 1.5 years cooling time by ORIGEN2 code. The shielding calculation for the cask has been made by using MCNP4A code with continuous cross section data library from ENDF/B-V. As a result of neutron dose rate analysis, another shielding calculational model on spent fuel shipping cask was provided which is using the Monte Carlo method.

• Atmosphere
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• v.29 no.5
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• pp.537-550
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• 2019

To analyze the observation environment of solar radiation stations operated by the Korea Meteorological Administration (KMA), we analyzed the skyline, Sky View Factor (SVF), and solar radiation due to the surrounding topography and artificial structures using a Digital Elevation Model (DEM), 3D camera, and solar radiation model. Solar energy shielding of 25 km around the station was analyzed using 10 m resolution DEM data and the skyline elevation and SVF were analyzed by the surrounding environment using the image captured by the 3D camera. The solar radiation model was used to assess the contribution of the environment to solar radiation. Because the skyline elevation retrieved from the DEM is different from the actual environment, it is compared with the results obtained from the 3D camera. From the skyline and SVF calculations, it was observed that some stations were shielded by the surrounding environment at sunrise and sunset. The topographic effect of 3D camera is therefore more than 20 times higher than that of DEM throughout the year for monthly accumulated solar radiation. Due to relatively low solar radiation in winter, the solar radiation shielding is large in winter. Also, for the annual accumulated solar radiation, the difference of the global solar radiation calculated using the 3D camera was 176.70 MJ (solar radiation with 7 days; suppose daily accumulated solar radiation 26 MJ) on an average and a maximum of 439.90 MJ (solar radiation with 17.5 days).

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.446.2-446.2
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• 2003

• Nuclear Engineering and Technology
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• v.2 no.3
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• pp.149-161
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• 1970

This paper describes a detailed investigational performance on the gamma radiation shield effect of heavy concretes that were manufactured by the use of mineral ores produced domestically and which may be possibly applied for the biological shield design. Ten different kinds of mineral ores were collected for use as the aggregates, physical test and chemical analysis for them were carried out to select the aggregate with a better property. Through the experimental investigation on the shielding effect of various concretes with different combination of concrete components such as water-cement and fine-coarse aggregate ratios, it was possible to derive some criteria for the best condition being capable of obtaining the concretes with high density and good uniformity. Data on the shielding-effectiveness of the different concretes were obtained by performing collimated beam experiment using 60Co gamma-ray. Analyzing the shielding-efficiency, shielding-concrete specific gravity and biological shield cost, the optimum condition of yielding the best economic shielding design, with low cost and good spatial distribution to some extent was determined.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.480-481
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• 2018

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.310.2-310
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• 2002

• The Journal of the Institute of Internet, Broadcasting and Communication
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• v.19 no.3
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• pp.59-65
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• 2019

The EMI shielding effectiveness of the shielding layer thickness was analyzed when the metal shielding layer was placed in the near field of the magnetic probe and the noise source. Microstrip lines were used as noise source, and graphite and ferrite were selected as metal shielding materials. The magnetic probe uses the electromagnetic radiation measurement method using the magnetic probe by applying the IEC 61967-6 method. The transmission coefficient between the microstrip line and the magnetic probe was analyzed. The distance between the two was 1 mm for a single shielding layer and 5 mm for a multiple shielding layer. The thickness of the shielding layer was changed to 5 um, 10 um, 30 um, and 50 um. When the frequency was changed from 150 kHz to 1 GHz, a maximum shielding effectiveness (SE) of 44.9 dB was obtained.

• Journal of Radiation Protection and Research
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• v.43 no.4
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• pp.154-159
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• 2018

Background: To prevent small leakage accidents, a real-time and direct detection system for small leaks with a detection limit below that of existing systems, e.g. $0.5gpm{\cdot}hr^{-1}$, is required. In this study, a small-size beta detector, which can be installed inside the reactor containment (CT) building and detect small leaks directly, was suggested and its feasibility was evaluated using MCNPX simulation. Materials and Methods: A target nuclide was selected through analysis of radiation from radionuclides in the reactor coolant system (RCS) and the spectrum was obtained via a silicon detector simulated in MCNPX. A window was designed to reduce the background signal caused by other nuclides. The sensitivity of the detector was also estimated, and its shielding designed for installation inside the reactor CT. Results and Discussion: The beta and gamma spectrum of the silicon detector showed a negligible gamma signal but it also contained an undesired peak at 0.22 MeV due to other nuclides, not the $^{16}N$ target nuclide. Window to remove the peak was derived as 0.4 mm for beryllium. The sensitivity of silicon beta detector with a beryllium window of 1.7 mm thickness was derived as $5.172{\times}10^{-6}{\mu}Ci{\cdot}cc^{-1}$. In addition, the specification of the shielding was evaluated through simulations, and the results showed that the integrity of the silicon detector can be maintained with lead shielding of 3 cm (<15 kg). This is a very small amount compared to the specifications of the lead shielding (600 kg) required for installation of $^{16}N$ gamma detector in inside reactor CT, it was determined that beta detector would have a distinct advantage in terms of miniaturization. Conclusion: The feasibility of the beta detector was evaluated for installation inside the reactor CT to detect small leaks below $0.5gpm{\cdot}hr^{-1}$. In future, the design will be optimized on specific data.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.957-962
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• 2018

The gamma radiation shielding capability (GRSC) of two clay-materials (Ball clay and Kaolin)of Southwestern Nigeria ($7.49^{\circ}N$, $4.55^{\circ}E$) have been investigated by determine theoretically and experimentally the mass attenuation coefficient, ${\mu}/{\rho}(cm^2g^{-1})$ of the clay materials at photon energies of 609.31, 1120.29, 1173.20, 1238.11, 1332.50 and 1764.49 keV emitted from $^{214}Bi$ ore and $^{60}Co$ point source. The mass attenuation coefficients were theoretically evaluated using the elemental compositions of the clay-materials obtained by Particle-Induced X-ray Emission (PIXE) elemental analysis technique as input data for WinXCom software. While gamma ray transmission experiment using Hyper Pure Germanium (HPGe) spectrometer detector to experimentally determine the mass attenuation coefficients, ${\mu}/{\rho}(cm^2g^{-1})$ of the samples. The experimental results are in good agreement with the theoretical calculations of WinXCom software. Linear attenuation coefficient (${\mu}$), half value layer (HVL) and mean free path (MFP) were also evaluated using the obtained ${\mu}/{\rho}$ values for the investigated samples. The GRSC of the selected clay-materials have been compared with other studied shielding materials. The cognizance of various factors such as availability, thermo-chemical stability and water retaining ability by the clay-samples can be analyzed for efficacy of the material for their GRSC.

• Journal of the Korean Society of Radiology
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• v.15 no.7
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• pp.957-964
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• 2021

Fluorine-18 used in PET/CT scans is a radioactive isotope that emits positrons, and high energy annihilation gamma rays and beta rays cause exposure to radiation workers. In this study, as part of a plan to reduce the exposure dose of radiation workers working in the Department of Nuclear Medicine, the cause of the low shielding efficiency of Apron for F-18 was identified, and the effectiveness of the Apron double-shielded with acrylic was evaluated. L-Block, Apron+acrylic, Apron, Acrylic+Apron, and Acrylic five shields are used to measure the dose, and the tendencies were compared by performing a Monte Carlo simulation. As a result, it was found that the shielding rate of Apron double shielded with acrylic was about 4 to 8% higher than that of Apron single shielded. To the extent that it does not significantly affect the user's activity, double-shielded personal protective clothing with an appropriate acrylic thickness could help reduce radiation workers' exposure.