• Journal of the Korea Institute of Information and Communication Engineering
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• v.23 no.7
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• pp.779-785
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• 2019

By applying an ergonomically developed Bismuth shield to the endovascular treatment of cerebral aneurysm the radiation dose of the scalp and lens from the medical radiation exposure was reduced. The enrtance surface dose was analyzed by measuring the occipital parts, bilateral temporal parts, bilateral quadriceps, and nasal tip of the developed bismuth shield using a photostimulable fluorescence dosimeter before (Group A) before use (Group B). Signal to noise ratio (SNR) and contrast to noise ratio (CNR) analysis were used to evaluate the image quality when Bismuth shielding was used. The mean entrance surface dose of A group and B group was 26.92% lower than that of A group. The analysis of CNR and SNR was the same for both Roadmap and DSA. The use of Bismuth shielding is an alternative that can reduce the radiation impairment due to temporary hair loss and other stochastic effects that may occur after cerebrovascular intervention.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4329-4334
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• 2023

In this research, a prompt gamma neutron activation analysis (PGNAA) system is designed and constructed based on the use of a low power research reactor. For this purpose, despite the fact that this reactor did not include beam tubes, a thermal neutron beam line is installed inside the reactor tank. The extraction of the beam line from inside the tank made it possible to provide the neutron flux from the order of 106 n.cm^-2.s^-1. Also, because the beam line is installed in a tangential position to the reactor core, its gamma level has been minimized. Also, a suitable radiation shield is considered for the detector to minimize the background radiation and prevent radiation damage to the detector. Calculations and measurements are done in order to characterize this system, as well as spectrometry of several samples. The results of evaluations and experiments show that this system is suitable for performing PGNAA.

• The Journal of the Korea Contents Association
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• v.15 no.7
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• pp.282-288
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• 2015

A monte carlo simulation about shielding material and thickness of the syringe shield for radiation shield was performed. As a result of analysis, high atomic number materials such as tungsten, lead and bismuth have the highest shielding effect. However, $^{18}F$, $^{67}Ga$ and $^{111}In$ show high energy distribution in the region with thin shielding thickness. As the thickness of shielding materials increased, the energy distribution decreased due to reduction of ${\gamma}$-ray. In the case of low atomic number materials, they, showed energy distribution from highest to lowest, were barium sulfate, steel, stainless, iron and copper. Aluminum, plastic, concrete and water showed diverse aspect. they showed relatively high energy distribution because of increased ${\gamma}$-ray that penetrate the shield.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.457-463
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• 2005

KHNP(Korea Hydro and Nuclear Power Ltd., Co.) is developing a HIC transport package which is satisfying domestic and IAEA regulations and NETEC(Nuclear Environment Technology Institute) is conducting a conceptual design. In this study, the shielding thickness was calculated using the data from radionuclide assay program which is currently using in nuclear sites and Micro Shield code. Considering the structural safety, carbon steel was chosen as shielding material and the shielding thickness was calculated for 500 R/hr and 100 R/hr at HIC surface, respectively. Through the shielding analysis, it was evaluated that the regulation limit is satisfied when the shielding thickness is 22 cm for 500 R/hr and 17 cm for 100/hr.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.419.2-419
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• 2001

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4137-4141
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• 2021

As NPPs continue to operate, liquid waste continues to be generated, and containers are needed to store and transport them at low cost and high capacity. To transport and store liquid phase very low-level radioactive waste (VLLW), a container is designed by considering related regulations. The design was constructed based on the existing container design, which easily transports and stores liquid waste. The radiation shielding calculation was performed according to the composition change of barium sulfate (BaSO₄) using the Monte Carlo N-Particle (MCNP) code. High-density polyethylene (HDPE) without mixing the additional BaSO₄, represented the maximum dose of 1.03 mSv/hr (<2 mSv/hr) and 0.048 mSv/hr (<0.1 mSv/hr) at the surface of the inner container and at 2 m away from the surface, respectively, for a 10 Bq/g of ⁶⁰Co source. It was confirmed that the dose from the inner container with the VLLW content satisfied the domestic dose standard both on the surface of the container and 2 m from the surface. Although it satisfies the dose standard without adding BaSO₄, a shielding material, the inner container was designed with BaSO₄ added to increase radiation safety.

• Nuclear Engineering and Technology
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• v.56 no.8
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• pp.3236-3241
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• 2024

Ion cyclotron resonance heating (ICRH) is an important auxiliary heating method applied to the China Fusion Engineering Test Reactor, which can effectively heat the ions and electrons in plasma. Owing to the harsh nuclear environment, neutronic analyses are required to verify tritium self-sufficiency and neutron-shielding requirements. In this study, a neutronics analysis of the ICRH antenna was conducted using the COre and System integrated engine for Reactor Monte Carlo (cosRMC) code to estimate the neutron flux, radiation damage, nuclear heating, gas generation rate of key components, and tritium breeding ratio (TBR), providing data support for the subsequent optimization of the shielding design. In addition, the neutron flux of the coils around the antenna was calculated to prevent the entry of neutrons that damage the magnetic field coils through the gaps between the port plugs and antenna, and the shielding effects of the port-plug antenna on the surrounding components were analyzed. Finally, the results obtained using the cosRMC and MCNP codes were compared, which and presented good agreement, thus verifying the reliability of the neutronic analysis using the cosRMC code.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.12
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• pp.1567-1572
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• 2013

To supply electric power in certain extreme environments such as a spacecraft or in military applications, a radioisotope thermoelectric generator has been highlighted as a useful energy source owing to its high energy density, long lifetime, and high reliability. A radioisotope thermoelectric generator generates electric power by using the heat energy converted from the radioactive energy of a radioisotope. In this study, FE analyses such as radiation shield analysis, heat transfer analysis, and power recovery rate analysis have been carried out to achieve an optimal design for a radioisotope thermoelectric generator using $SrTiO_2$.

• Journal of Radiation Industry
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• v.9 no.2
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• pp.69-75
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• 2015

Purpose: The gonad shielding is used to minimize the impact of the exposure to gonads when Abdomen, Pevis and Hip X-ray inspections are conducted on radiation impressionable pediatric patients. By the way, the gonad is palpable difficult and impossible to check visually because it's a sensitive area, so tests are conducted with the approximate location of shielding, thereby appearing problems of not shielding gonads accurately. Accordingly, this study produced shields by age and gender of pediatric patients and studied the method of positioning shields with ASIS as a reference point without palpable sensitive areas, and tried to evaluate its usability. Materials and methods: The study surveyed 30 pediatric patients by gender and age, who came and got inspected in Department of Radiology, our hospital from February 2012 to January 2014 and obtained the value of tolerance by measuring the average size of the pelvis using the distance measurement function of Infinitt Piview with the images stored in the PACS and producing shields by age and gender of pediatric patients and specifying the areas at random for the comparative analysis of pre- and post-using. It calculated the technology statistics($mean{\pm}SD$) with the value of tolerance measured the length using SPSS 12.0 statistical program. Results: As for boys, differences in the tolerance range of pre- and post-using shields were 2.69 mm in case of 1 year old, 2.58 mm in 2 years, 2.37 mm in 3 years, 2.815 mm in 4~5 years, 2.043 mm in 7~10 years, and as for girls, 1.92 mm in 1~2 years, 1.75 mm in 3~4 years, 2.52 mm in 5~6 years and 1.93 mm in 7~10. After analyzing the pre- and post-using shields for all of boys and girls, there were statistically significant differences(P<0.050). Conclusions: It is considered that we can minimize the exposure to gonads and get a better video for diagnosis in testing high biological impressionable pediatric, if we use shields correctly with ASIS as a reference point considering its shape and size by age and gender in Abdomen, Pevis and Hip X-ray inspections.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2740-2747
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• 2024

The sintering process acquired ferromagnetic copper ferrite ceramic material with a small concentration of Ni ion at 1100 ℃ for 1 h. Previously, copper ferrite with Ni proportions powder was acquired by the wet chemical process according to the relation CuFe_2-xNi_xO₄ where x takes values 0.0, 0.015, 0.03, 0.04, and 0.05. The role of Ni ion in the copper ferrite structure was investigated by X-ray analysis, Scanning electron microscope, EDX analysis, and density measurements. The gamma-ray shielding properties for the fabricated CuFeNiO ceramics samples were evaluated using the Monte Carlo simulation method. The obtained results show an enhancement in the linear attenuation coefficient for the fabricated ceramics with increasing the insertions of Ni ions within the fabricated samples, where increasing the Ni ions concentration between 0 and 1.19 wt% increases the linear attenuation by between 1.581 and 1.771 cm^-1 (at 0.103 MeV), 0.304-0.338 cm^-1 (at 0.662 MeV), and 0.160-0.178 cm^-1 (at 2.506 MeV), respectively. Simultaneously, the radiation protection efficiency for a 1 cm thickness of the fabricated samples increased between 14.8 and 16.3% with increasing the Ni ions between 0 and 1.19 wt%. Although the Ni doping concentration does not exceed 1.5 wt% of the total composition of the fabricated ceramics, the shielding capacity of the fabricated ceramics was enhanced by more than 11%, along the studied energy interval. Therefore, the fabricated samples can be used in gamma-ray shielding applications.