• Journal of the Korean Society of Radiology
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• v.7 no.6
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• pp.403-408
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• 2013

This study made a tentative estimation of the shielding rate of barium compound by thickness through monte carlo simulation to apply medical radiation shielding products that can replace existing lead. Barium sulfate($BaSO_4$) was used for the shielding material, and thickness of the shielding material specimen was simulated from 0.1 mm to 5 mm by applying $15{\times}15cm^2$ of specimen area, $4.5g/cm^3$ of density of barium sulfate, and $11.34g/cm^3$ density of lead. Entered source was simulated with 10kVp Step in consecutive X-ray energy spectrum(40 kVp ~ 120 kVp). Absorption probability in 40 kVp ~ 60 kVp showed same shielding rate with lead in 3 mm ~ 5 mm of thickness, but it was identified that under 2 mm, the shielding rate was a bit lower than the existing lead shielding material. Also, the shielding rate in 70 kVp ~ 120 kVp energy band showed similar performance as the existing lead shielding material, but it was tentatively estimated as fairly low shielding rate below 0.5 mm. This study estimated the shielding rate of barium compound as the thickness function of x-ray energy band for medical radiation through monte carlo simulation, and made comparative analysis with existing lead. Also, this study intended to verify application validity of the x-ray shielding material for medical radiation of pure barium sulfate. As a result, it was estimated that the shielding effect was 95% higher than the existing lead 1.5 mm in at least 2 mm thickness of barium compound in medical radiation energy band 70 kVp ~ 120 kVp, and this result is considered valid to be provided as a base data in weight lightening production of radiation shielding product for medical radiation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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• pp.349-356
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• 2022

Nuclear power plant decommissioning generates significant concrete waste, which is slightly contaminated, and expected to be classified as clearance concrete waste. Clearance concrete waste is generally crushed into rubble at the site or a satellite treatment facility for practical disposal purposes. During the process, workers are exposed to radiation from the nuclides in concrete waste. The treatment processes consist of concrete cutting/crushing, transportation, and loading/unloading. Workers' radiation exposure during the process was systematically studied. A shielding package comprising a cylindrical and hexahedron structure was considered to reduce workers' radiation exposure, and improved the treatment process's efficiency. The shielding package's effect on workers' radiation exposure during the cutting and crushing process was also studied. The calculated annual radiation exposure of concrete treatment workers was below 1 mSv, which is the annual radiation exposure limit for members of the public. It was also found that workers involved in cutting and crushing were exposed the most.

• Journal of Radiation Protection and Research
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• v.32 no.4
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• pp.190-193
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• 2007

The objective of this study is to recommend the radiation protection design parameters from the shielding point of view for concrete wall between the decay tank room and the primary pump room in TRIGA Mark-II Research Reactor Facility. The shield design for this concrete wall has been performed with the help of Point-kernel Shielding Code Micro-Shield 5.05 and this design was also validated based on the measured dose rate values with Radiation Survey Meter (G-M Counter) considering the ICRP-60 (1990) recommendations for occupational dose rate limit ($10{\mu}Sv/hr$). The recommended shield design parameters are: (i) thickness of 114.3 cm Ilmenite-Magnetite Concrete (IMC) or 129.54 cm Ordinary Reinforced Concrete (ORC) for concrete wall A (ii) thickness of 66.04 cm Ilmenite-Magnetite Concrete (IMC) or 78.74 cm Ordinary Reinforced Concrete (ORC) for concrete wall B and (iii) door thickness of 3.175 cm Mild Steel (MS) on the entrance of decay tank room. In shielding efficiency analysis, the use of I-M concrete in the design of this concrete wall shows that it reduced the dose rate by a factor of at least 3.52 times approximately compared to ordinary reinforced concrete.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2444-2451
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• 2024

This work demonstrates a new sol-gel approach for synthesizing PbO₂-doped zirconia using zircon mineral precursors. The streamlined methodology enables straightforward fabrication of the doped zirconia composites. Comprehensive materials characterization was performed using XRD, SEM, and TEM techniques to analyze the crystal structure, microstructure, and morphology. Quantitative analysis of the XRD data provided insights into the nanoscale crystallite sizes achieved, along with their relationship to lattice imperfections. Furthermore, the gamma-ray shielding capacity for the PbO₂-doped zirconia samples was estimated by the Monte Carlo simulation, which proves an increase in the gamma ray shielding properties by raising the Pb concentration. The linear attenuation coefficient increased between 0.467 and 0.499 cm^-1 (at 0.662 MeV) by increasing the Pb content between 11 and 21 wt%. By increasing the Pb content to 21 wt%, the synthesized composites' lead equivalent thickness reaches 2.49 cm. The radiation shielding properties for the synthesized composites revealed a remarkable performance against low and intermediate γ-ray photons, with radiation shielding capacity of 37.3 % and 21.4 % at 0.662 MeV and 2.506 MeV, respectively. As a result, the developed composites can be employed as an alternative shielding material in hospitals and radioactive zones.

• Journal of Radiation Protection and Research
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• v.12 no.1
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• pp.34-39
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• 1987

Radiation shielding analysis for a 6MeV X-ray facility was carried out. The primary and leakage radiation for the facility can be evaluated based on the methodology in NCRP No. 49 and 51. The present study deals with radiation scattering analysis for the outside and inside door of the facility based on the albedo concept. The calculated dose rates were compared with the results of MORSE-CG code calculation and the measured data, resulting in a good agreement, even though there existed some deviation for the inside door. These results can be utilized to the radiation shielding design of the medical and industrial X and gamma ray facilities, and to the safety evaluation of these facilities.

• Journal of Radiation Protection and Research
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• v.22 no.2
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• pp.77-83
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• 1997

Epoxy resin-type neutron shielding materials, KNS(Kaeri Neutron Shield)-101, KNS-102, and KNS-103 have been fabricated to be used in spent fuel shipping cask. The base material is epoxy resin, and polypropylene, aluminium hydroxide, and boron carbide are added. These shielding materials offer good fluidity at processing, which makes it possible to apply this resin shield to complicated geometric shapes such as shipping cask. The shielding property of these shielding materials for shipping cask for loading 28 PWR spent fuel assemblies has been evaluated. ANISN code is used to evaluate the shielding property of the shipping cask with the thickness of the three neutron shielding materials greater than 10 cm. As a result of analysis, the maximum calculated dose rate at the radial surface of the cask is determined to be $300{\mu}Sv/h$ and the maximum calculated dose rate at 100 cm from the cask is $97{\mu}Sv/h$. These dose rates remain within allowable values specified in related regulations.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1783-1790
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• 2023

Barium Bismuth Oxide Borate (BBOB) has been synthesized for the first time using solution combustion technique. SEM analysis reveal flower shape of the nanoparticles. The formation of the nanoparticles has been confirmed through XRD & FTIR studies which gives the physical and chemical structure of the novel material. The UV light absorption is observed in the range 200-300 nm. The present study highlights the radiation shielding ability of BBOB for different radiations like X/Gamma rays, Bremsstrauhlung and neutrons. The gamma shielding efficiency is comparable to that of lead in lower energy range and lesser than lead in the higher energy range. The bremsstrauhlung exposure constant is comparably larger for BBOB NPs than that of concrete and steel however it is lesser than that of lead. The beauty of BBOB nanoparticles lies in, high absorption of radiations and low emission of secondary radiations when compared to lead. In addition, the neutron shielding parameters like scattering length, absorption and scattering cross sections of BBOB are found to be much better than lead, steel and concrete. Thus, BBOB nanoparticles are highly efficient in absorbing X/Gamma rays, neutrons and bremsstrauhlung radiations.

• Journal of Radiation Protection and Research
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• v.41 no.2
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• pp.141-147
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• 2016

Background: A project for building a neutron time-of-flight (nTOF) facility is progressing. We expect that the construction will start in early 2016. Before that, a detailed simulation based on the current architectural drawings was performed to optimize the performance of our facility. Materials and Methods: Currently, several parts had been modified or changed from the original design to reflect requirements such as the layout of the electron beam line, shape of the vacuum chamber producing a neutron beam, and the underground layout of the nTOF facility. Detailed analysis for these modifications has been done with MCNP simulation. Results and Discussion: An overview of our photo-neutron source and KAERI nTOF facility were introduced. The numerical simulations for heat deposition, source term, and radiation shielding of KAERI nTOF facility were performed and the results are discussed. Conclusion: We are expecting that the construction of the KAERI nTOF facility will start in early 2016, and these results will be used as basic data.

• Progress in Medical Physics
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• v.29 no.4
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• pp.115-122
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• 2018

Although the current internationally recommended standard for the use factor (U) applied to CyberKnife is 0.05 (5%), the CyberKnife shielding standard is applied more stringently. This study, based on clinical data, was aimed at examining the appropriateness of existing shielding guidelines. Sixty patients treated with G4 CyberKnife were selected. The patients were divided into two groups, according to whether they underwent skull or spine tracking. Based on the results, the use factors for each wall ranged from 0.028 (2.8%) to 0.031 (3.1%) for the intracranial treatment and 0.020 (2.0%) to 0.022 (2.2%) for the body treatment. Excessive barrier thickness resulted in inefficient use of space and higher cost to the institutions. Furthermore, because the use factor is influenced by the position of the robot, the use factor determined based on the clinical data of this study would facilitate more reasonable treatment room design.

• Journal of Veterinary Clinics
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• v.37 no.1
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• pp.15-22
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• 2020

This study investigated the effectiveness of radiation safety rules in animal hospital and the awareness and behavior of veterinary radiation workers. With the questionnaires, the data was collected from randomly selected veterinarians in animal hospitals and animal medical imaging centers. Collected data were about radiation device, shielding device, regulations, safety management, education, knowledge, behavior and awareness. Frequency, correlation and multiple regression analysis were performed. The medical devices related with radiation in animal hospital were X-ray (59%), CT (15%), fluoroscopy (12%), mobile X-ray (12%) and others (2%). The number of people using radiation shielding device is high. The answers were low on knowing radiation related regulation and receiving radiation protection education. The group with higher knowledge and awareness shows positive correlation with safety behavior. The increase of use of the radiation related medical devices in veterinary hospital causes the increase of radiation exposure risk. This study suggests that radiation safety management system and policies need to be developed to protect radiation workers and give them correct information and consciousness.