• Nuclear Engineering and Technology
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• v.56 no.8
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• pp.3335-3346
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• 2024

Due to the radiation released by commonly used isotopes, many nuclear, medical, and industrial facilities require proper radiation shielding. In this work, distinct copper borate glasses intercalated with varied aluminum and vanadium oxide (Al₂O₃ and V₂O₅) content have been synthesized and used against radiation (gamma rays and fast/thermal neutrons). The different percents were as follows: [60% B₂O₃ + 35% CuO + (5-x)% Al₂O₃ + xV₂O₅], where x = 0, 1, and 2.5 wt.%, which was coded as BCu(5-x)Al:xV. The synthesized glass samples were characterized using Fourier transforms, infrared, and X-Raydiffraction analysis. Experimentally, the radiation shielding possessions of the samples were established using an HPGe detector at the gamma energy lines 0.356 MeV, 0.661 MeV, 1.173 MeV, and 1.332 MeV. Also, the prepared glasses were investigated theoretically using the Monte Carlo code (MCNP5) at photon energies of 0.015-15 MeV. Also, the fast and thermal neutron macroscopic effective removal cross-sections were calculated using MRCsC and JANIS-4.1 software, respectively. The prepared sample BCu2.5Al:2.5V, which has a vanadium and aluminum content of 2.5%, has the highest linear attenuation coefficient as well as the highest removal cross-section for fast, and thermal neutrons.

• 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.

• Journal of the Korean Society of Radiology
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• v.17 no.5
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• pp.641-649
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• 2023

This study compares and analyzes the performance of a shield manufactured using 3D printing technology to find out its applicability as a shield in high-energy electron beam therapy. Actual measurement and monte carlo simulations were performed to evaluate the shielding performance of 3D printing materials for high-energy electron beams. First, in order to secure reliability for the simulation, a source term evaluation was conducted by referring to the IAEA's TRS-398 recommendation. Second, to analyze the shielding performance of PLA+W (93%), a specimen was manufactured using a 3D printer, and the shielding rate by thickness according to electron beam energy was evaluated. Third, the shielding thickness required for electron beam treatment was calculated through a comparative analysis of shielding performance between PLA+W (93%) and existing shielding bodies. First, as a result of the evaluation of the source term through actual measurement and simulation, the TRS-398 recommendation was satisfied with an error of less than 1%, thereby securing the reliability of the simulation. Second, as a result of the shielding performance analysis for PLA+W (93%), 6 MeV electron beams showed a shielding rate of more than 95% at 3.12 mm, and 15 MeV electron beams showed a shielding rate of more than 90% at 10 mm thickness. Third, through simulations, comparative analysis between PLA+W (93%) materials and existing shields showed high shielding rates within the same thickness in the order of tungsten, lead, copper, PLA+W (93%), and aluminum. 6 MeV electron beams showed almost similar shielding rates at 5 mm or more and 15 MeV electron beams. Through this study in the future, it is judged that it can be used as basic data for the production and application of shielding bodies using PLA+W (93%) materials in high-energy electron beam treatment.

• Radiation Oncology Journal
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• v.12 no.2
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• pp.253-262
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• 1994

The patterns of the conventional radiation treatment fields and their shielding blocks are analysed to determine the optimal dimension of the MultiLeaf Collimator (MLC) which is considered as an essential tool for conformal therapy. Total 1109 radiation fields from 303 patients (203 from Asan Medical center, 50 from Baek Hosp and 50 from Hanyang Univ. Hosp.) were analysed for this study. Weighted case selection treatment site (from The Korean Society of Therapeutic Radiology 1993). Ninety one percent of total fields have shielding blocks. Y axis is defined as leaf movement direction and it is assumed that MLC is installed on the cranial-caudal direction. The length of X axis were distributed from 4cm to 40cm (less than 21cm for $95\%$ of cases), and Y axis from 5cm to 38cm (less than 22cm for $95\%$ of cases). The shielding blocks extended to less than 6cm from center of the field for $95\%$ of the cases. Start length for ninety five percent of block is less than 10cm for X axis and 11cm for Y axis. Seventy six percent of shielding blocks could be placed by either X or Y axis direction, $7.9\%$ only by Y axis, $5.1\%$ only by X axis and It is reasonable to install MLC for Y direction. Ninety five percent of patients can be treated with coplanar rotation therapy without changing the collimator angle. Eleven percent of cases of cases were impossible to replace with MLC. Futher study of shielding technique is needed for $11\%$ impossible cases. The treatment field dimension of MLC should be larger than $21cm{\times}22cm$. The MLC should be designed as a pair of 21 leaves with 1cm wide for an acceptable resolution and 17cm long to enable the leaf to overtravel at least 6cm from the treatment field center.

• Journal of Electrical Engineering and Technology
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• v.13 no.2
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• pp.936-942
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• 2018

The diagnostic multi-leaf collimator preventing unnecessary dose from entering into patients during the diagnostic examination was made in this study. The movement of the entire 50 leaves was embodied with the group of 25 ones thereof configured in a pair facing each other on the left and right of the median line. Dimensions of the length, width, and height of each shielding leaf were $5{\times}0.5{\times}0.5cm^3$ resulting in the maximum boost field of $10{\times}10cm^2$. The material of multi-leaf collimator had the excellence on the machinability with the use of the SKD-11 alloy tool steel having the high wear resistance against frequent movement, and it was devised to control both-side's shielding leaves by moving 2 motors unlike existing remedial multi-leaf collimator that use as many motors as the number of 50 shielding leaves. Thereafter, the transmission dose of leaves, cross-leaf leakage dose, and inter-leaf leakage dose were measured by the developed multi-leaf collimator attached to X-ray equipment. An ionization chamber was used to detect doses there from, and the comparative analysis was carried out by means of the radiographic film that was easy to detect the dose leakage in between each leaf. Results obtained from the test conducted in comparative analysis yielded approximately 98%, 96%, and 94% of shielding efficiency realized at each level of energy of 80kV, 100kV, and 120kV it was confirmed there was no dose leakage resulted from the varied level of irradiation energy. Thus the multi-leaf collimator to be developed based on this study is thought that it could fully reduce the unnecessary dose to patients in the diagnostic test and the shielding efficiency thereof is expected to be increasing if it is made in a miniaturized form with a way of increasing the thickness of each leaf later for an extended application to general diagnostic purposes.

• Journal of the Korean Society of Radiology
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• v.17 no.2
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• pp.215-221
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• 2023

Among brain CT scan conditions including the lens, the tube voltage was changed to 80, 100, and 120 kVp and applied. The change in dose was analyzed using lead, lead goggles and barium sulfate silicon shielding materials, and the degree of influence of the shielding materials on image quality was compared and analyzed by applying the SNR, CNR, and SSIM index analysis methods. As a result, it was analyzed that although the dose was reduced by applying all shielding materials, the difference in dose reduction was not large (P > 0.05). In addition, as for the change in image quality due to the application of the shielding material, SNR and CNR were the highest when lead goggles were applied, and the structural similarity was measured to be the best as it was closest to the reference value of 1 in SSIM analysis. Therefore, based on the results of this study, it is thought that if more diverse shielding materials and clinical test results are derived and applied, it will be helpful for the clinical application criteria in the case of shielding utilization inspection.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2153-2162
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• 2024

The purpose of this study is to develop a high-energy neutron shielding material applied in proton therapy environment. Composite shielding material consisting of 10.00 wt% boron carbide particles (B₄C), 13.64 wt% surface-modified cross-linked polyethylene (PE), and 76.36 wt% tungsten particles were fabricated by hot-pressure sintering method, where the optimal ratio of the composite is determined by the shielding effect under the neutron field generated in typical proton therapy environment. The results of Differential Scanning Calorimetry measurements (DSC) and tensile experiment show that the composite has good thermal and mechanical properties. In addition, the high energy-neutron shielding performance of the developed material was evaluated using cyclotron proton accelerator with 100 MeV proton. The simulation shows a 99.99% decrease in fast neutron injection after 44 cm shielding, and the experiment result show a 99.70% decrease. Finally, the shielding effect of replacing part of the shielding material of the proton therapy hall with the developed material was simulated, and the results showed that the total neutron injection decreased to 0.99‰ and the neutron dose reduced to 1.10‰ before the enhanced shielding. In summary, the developed material is expected to serve as a shielding enhancement material in the proton therapy environment.