• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.25 no.6
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• pp.57-67
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• 2011

Purpose of this study is to find the mitigation method of magnetic field by finite length multi-conductors such as indoor distribution lines and to be applicable to design of the distribution lines. For this purpose, exact formula about the components $B_x$, $B_y$, $B_z$ of magnetic field need in case of straight line-conductor with finite length forward any direction. In this study simple formula of the components were deduced and by using these formula magnetic fields for various models of line-configurations were calculated. And also a calculation method of induced currents in conductive shield was presented and using this method, programing of calculation is relatively easy and calculation time is short. The magnetic field after cancellation by these induced currents was calculated. All of calculations were performed by Matlab 7.0 programs. Through the calculation results it could be obtained followings for the mitigation of magnetic fields. The separation between conductors ought to be smaller than smaller as possible. In case of 3-phase, delta configuration is more effective than flat configuration. In case of 3-phase, unbalanced currents ought to be reduced as possible.. In case of more than two circuits of 3-phase, adequate locations of each phase-conductor such as rotating configuration of 3-phase conductors are more effective. The magnetic shielding effect of the conductive shielding sheet is very high.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2016.05a
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• pp.773-774
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• 2016

In this study, we perform the weight reduction and miniaturization of the shielding element that is applied for gamma-ray detectors for imaging of gamma-ray source. Through previous studies, we implemented a lead-based shielding element that represents the shielding effectiveness and performance of commercially available gamma-ray imaging apparatus similar to the shielding body. In this paper, we designed a tungsten-based shield for weight reduction and miniaturization than lead-based shield. We performed the MCNP simulation for shield design and then we obtained the results of reducing the weight of the 17% and 51% of the volume.

• Progress in Medical Physics
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• v.32 no.4
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• pp.83-91
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• 2021

Various types of high-precision radiotherapy, such as intensity-modulated radiation therapy (IMRT), tomotherapy (Tomo), and stereotactic body radiation therapy have been available since 1997. After being covered by insurance in 2015, the number of IMRT cases rapidly increased 18-fold from 2011 to 2018 in Korea. IMRT, which uses a high-beam irradiation monitor unit, requires higher shielding conditions than conventional radiation treatments. However, to date, research on the shielding of facilities using IMRT and the current understanding of its status are insufficient, and detailed safety regulation procedures have not been established. This study investigated the recommended criteria for the shielding evaluation of facilities using medical linear accelerators (LINACs), including 1) the current status of safety management regulations and systems in domestic and international facilities using medical LINACs and 2) the current status of the recommended standards for safety management in domestic and international facilities using medical LINACs. It is necessary to develop and introduce a safety management system for facilities using LINACs for clinical applications that is suitable for the domestic medical environment and corresponds to the safety management systems for LINACs used overseas.

• Journal of Radiation Protection and Research
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• v.30 no.1
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• pp.39-44
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• 2005

A new developing reactor isn't fixed the structure and the materials of reactor components. To perform the shielding analysis for a reactor vessel by $R-\theta$ geometry, it takes much effort and time to modeling of source term according to the change of reactor components every time. Therefore, we developed the shielding analysis system for the reactor vessel by $R-{\theta}$ geometry, which wasn't affected by the reactor core geometry. By using the developed shielding analysis system, we performed the shielding analysis for the reactor vessel of an integral reactor which has the hexagonal geometry of nuclear fuel assemblies in reactor core. We compared the results obtained from the developed system with those obtained from MCNP analysis. Because the results of developed shielding analysis system were more conservative than those of MCNP calculation, it is useful for shielding analysis. As we had developed the new shielding analysis system for a reactor vessel by $R-{\theta}$ geometry, we reduced error of model for reactor core which was formerly designed by hand and saved the time and the effort to design source term model of reactor core.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.524-531
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• 2003

The usefulness of vitrification technology for low- and intermediate- level radioactive wastes was demonstrated because of high volume reduction, mechanical and chemical stability of final waste forms. Thus, a construction of the commercial vitrification plant Is currently promoted. Due to the high radiation level of the waste, shielding analysis has to be carried out for safe design in a vitrification facility. Because there has been no experience in the construction and operation of the vitrification facility in Korea, in this study, in order to get some information for help the detailed design and plan for operation in vitrification facility, shielding analysis for each facility in pilot plant is carried out by using source term from established study. For the selection of the shielding material, concrete was chosen compared to the lead because of economic advantage, weight consideration, and thermal resistance.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3855-3863
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• 2022

An evaluation of the radiation shielding performance of high-Z-particle-loaded polylactic acid (PLA) composite materials was pursued. Specimens were produced via fused deposition modeling (FDM) using copper-PLA, steel-PLA, and BaSO₄-PLA composite filaments containing 82.7, 75.2, and 44.6 wt% particulate phase contents, respectively, and were tested under broad-band flash x-ray conditions at the Sandia National Laboratories HERMES III facility. The experimental results for the mass attenuation coefficients of the composites were found to be in good agreement with GEANT4 simulations carried out using the same exposure conditions and an atomistic mixture as a model for the composite materials. Further simulation studies, focusing on the Cu-PLA composite system, were used to explore a shield design parameter space (in this case, defined by Cu-particle loading and shield areal density) to assess performance under both high-energy photon and electron fluxes over an incident energy range of 0.5-15 MeV. Based on these results, a method is proposed that can assist in the visualization and isolation of shield parameter coordinate sets that optimize performance under targeted radiation characteristics (type, energy). For electron flux shielding, an empirical relationship was found between areal density (AD), electron energy (E), composition and performance. In cases where ${\frac{E}{AD}}{\geq}2MeV{\bullet}cm{\bullet}g^{-1}$, a shield composed of >85 wt% Cu results in optimal performance. In contrast, a shield composed of <10 wt% Cu is anticipated to perform best against electron irradiation when ${\frac{E}{AD}}<2MeV{\bullet}cm{\bullet}g^{-1}$.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.785-794
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• 2016

The Argonne National Laboratory of the United States and the Kharkov Institute of Physics and Technology of the Ukraine have been collaborating on the design, development and construction of a neutron source facility at Kharkov Institute of Physics and Technology utilizing an electron-accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100-MeV electrons. The facility was designed to perform basic and applied nuclear research, produce medical isotopes, and train nuclear specialists. The biological shield of the accelerator building was designed to reduce the biological dose to less than 5.0e-03 mSv/h during operation. The main source of the biological dose for the accelerator building is the photons and neutrons generated from different interactions of leaked electrons from the electron gun and the accelerator sections with the surrounding components and materials. The Monte Carlo N-particle extended code (MCNPX) was used for the shielding calculations because of its capability to perform electron-, photon-, and neutron-coupled transport simulations. The photon dose was tallied using the MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is very small, ~0.01 neutron for 100-MeV electron and even smaller for lower-energy electrons. This causes difficulties for the Monte Carlo analyses and consumes tremendous computation resources for tallying the neutron dose outside the shield boundary with an acceptable accuracy. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were utilized for this study. The generated neutrons were banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron dose. The weight windows variance reduction technique was also utilized for both neutron and photon dose calculations. Two shielding materials, heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total dose outside the shield boundary less than 5.0e-03 mSv/h during operation. The shield configuration and parameters of the accelerator building were determined and are presented in this paper.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2919-2926
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• 2023

In this paper, rare earth element (REE)/polyester composites were designed with lanthanum oxide, gadolinium oxide, and lutetium oxide as ray shielding agents, and polyethylene terephthalate (PET) as the base. Monte Carlo simulation was carried out using FLUKA software. We found that the radiation protection performance of the composite is affected by the type and amount of REE; a higher amount of REE equated to a better radiation protection performance of the composite. When the thickness of the composite and total thickness of the REE is constant, the number of superimposed layers inside the composite does not affect its shielding performance. Compared with a single-type REE/PET composite, a mixed-type REE/PET composite has a wider range of γ-ray absorption and better radiation protection performance. When the mass ratio of PET to REE is 2:8 and different types of REE are mixed with equal mass, several 0.2 cm-thick mixed-type REE/PET composites can shield >70% of 60 and 80 KeV γ-rays.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2000.11a
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• pp.628-632
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• 2000

In the recent day, shielding can was made of permalloy materials for automobile display instrument. The good roundness has an effect on magnetic property such as the low coercivity and the high permeability. But the products having the roundness is transformed by the heat treatment, the sending to the company and the assembly process. So for the stiffness of the edge, it is necessary for hemming process to be added in the deep drawing process. And it has the good appearance to create a smooth edge rather than a razor edge with urr. In this paper, it is controlled to get the best hemming product by the Pre-hemming angle($105^{\circ}$, $120^{\circ}$, $135^{\circ}$). And Possible process and tool design modification, which may lead to quality improvement in hemming were tested experimentally and using FEM. The commercial finite element program PAM-STAMPTM was used to simulate the pre-hemming process and hemming process, and the predictions were compared with experimental results according to the pre-hemming angle.

• Journal of Astronomy and Space Sciences
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• v.41 no.1
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• pp.17-23
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• 2024

The safety of electronic components used in aerospace systems against cosmic rays is one of the most important requirements in their design and construction (especially satellites). In this work, by calculating the dose caused by proton beams in geostationary Earth orbit (GEO) orbit using the MCNPX Monte Carlo code and the MULLASSIS code, the effect of different structures in the protection of cosmic rays has been evaluated. A multi-layer radiation shield composed of aluminum, water and polyethylene was designed and its performance was compared with shielding made of aluminum alone. The results show that the absorbed dose by the simulated protective layers has increased by 35.3% and 44.1% for two-layer (aluminum, polyethylene) and three-layer (aluminum, water, polyethylene) protection respectively, and it is effective in the protection of electronic components. In addition to that, by replacing the multi-layer shield instead of the conventional aluminum shield, the mass reduction percentage will be 38.88 and 39.69, respectively, for the two-layer and three-layer shield compared to the aluminum shield.