• Journal of radiological science and technology
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• v.41 no.6
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• pp.581-586
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• 2018

This study was conducted to analyze factors Influencing Protective Behavior against Radiation Exposure using questionnaires for 231 radiological technologists working in Computed Tomography(CT) examination room with high radiation dose in diagnostic radiology field. Statistical analysis of the collected data revealed that the reasons for partially shielding the examination part in the CT scan were the lack of protective equipment, securing of radiation justification, being annoying and maybe not being harm to adults in order. It was also revealed that the variables influencing the protective behavior were protective behavior against radiation harm, self-efficacy, protective environment, organization culture, protective knowledge and protective instrument in order. The higher the radiological protective environment(${\beta}=0.245$) and the lower the radiological protective knowledge(${\beta}=-0.034$), the more influential the protective behavior against radiation harm was. In this study, it was shown that non examination parts were not shielded in the CT scan. Therefore, it is necessary to improve the level of protective environment, to cultivate knowledge to improve the protective behavior against radiation harm and to have an intervention strategy for concrete action.

• 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.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.11 no.3
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• pp.1035-1039
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• 2010

Activation products generated by photoneutrons in concrete shielding wall around electron linac were estimated for a high energy X-ray container cargo inspection system. Monte carlo code, MCNPX2.5.0 was used for reference system of 9MeV fixed type dual-direction container cargo inspection system installed at major harbors in Korea. Activation products inventory generated by photoneutron (n,$\gamma$) reaction are estimated, and then radiation dose rate is calculated from the results.

• Nuclear Engineering and Technology
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• v.46 no.1
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• pp.73-80
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• 2014

A concrete cask is an option for spent nuclear fuel interim storage. A concrete cask usually consists of a metallic canister which confines the spent nuclear fuel assemblies and a concrete overpack. When the overpack undergoes a missile impact, which might be caused by a tornado or an aircraft crash, it should sustain an acceptable level of structural integrity so that its radiation shielding capability and the retrievability of the canister are maintained. A missile impact against a concrete overpack produces two damage modes, local damage and global damage. In conventional approaches [1], those two damage modes are decoupled and evaluated separately. The local damage of concrete is usually evaluated by empirical formulas, while the global damage is evaluated by finite element analysis. However, this decoupled approach may lead to a very conservative estimation of both damages. In this research, finite element analysis with material failure models and element erosion is applied to the evaluation of local and global damage of concrete overpacks under high speed missile impacts. Two types of concrete overpacks with different configurations are considered. The numerical simulation results are compared with test results, and it is shown that the finite element analysis predicts both local and global damage qualitatively well, but the quantitative accuracy of the results are highly dependent on the fine-tuning of material and failure parameters.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3317-3323
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• 2022

Half Value Layer calculations theoretically need prior specification of linear attenuation calculations, since the HVL value is derived by dividing ln(2) by the linear attenuation coefficient. The purpose of this study was to establish a direct computational model for determining HVL, a vital parameter in nuclear radiation safety studies and shielding material design. Accordingly, a typical gamma-ray transmission setup has been modeled using MCNPX (version 2.4.0) general-purpose Monte Carlo code. The MCNPX code's INPUT file was designed with two detection locations for primary and secondary gamma-rays, as well as attenuator material between those detectors. Next, Half Value Layer values of some well-known gamma-ray shielding materials such as lead and ordinary concrete have been calculated throughout a broad gamma-ray energy range. The outcomes were then compared to data from the National Institute of Standards and Technology. The Half Value Layer values obtained from MCNPX were reported to be highly compatible with the HVL values obtained from the NIST standard database. Our results indicate that the developed INPUT file may be utilized for direct computations of Half Value Layer values for nuclear safety assessments as well as medical radiation applications. In conclusion, advanced simulation methods such as the Monte Carlo code are very powerful and useful instruments that should be considered for daily radiation safety measures. The modeled MCNPX input file will be provided to the scientific community upon reasonable request.

• Journal of the Korea institute for structural maintenance and inspection
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• v.23 no.2
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• pp.107-115
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• 2019

The many countries are facing the shortage of natural resources, and the supply of aggregates are being exhausted. To consider this situation a variety of studies were performed for the development of alternative resources. In particular, high density filler material was used for shielding radioactive waste, large amount of natural aggregates are required in order to produce filler material. Also, in order to improve the shielding performance of filler material, it is required to increase the density of the filler material. Therefore, in this study was carried out to provide basic data for expanding the feasibility of high density industrial waste resource as aggregate in heavyweight concrete. From the test results, OPC case, concrete strength decreased by using heavyweight waste glass as fine aggregate, however, it is improved by using mineral admixture as binder. Therefore, when the heavyweight waste glass and steel slag are applied to heavyweight concrete, it is desirable to use mineral admixture, especially to use BFS than FA. Meanwhile, when the steel slag was replaced as coarse aggregate of heavyweight concrete, elasticity of modulus and radiation shielding performance can be improved owing to high density of steel slag.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.80-90
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• 1993

Radiation streaming through penetrations has been of great concern in radiation shielding design and analysis. This study developed a Monte Carlo method and constructed a data library of results calculated by the Monte Carlo method for radiation streaming through a straight cylindrical duct in concrete walls of a broad, mono-directional, mono-energetic gamma-ray beam of unit intensity. It was demonstrated that average dose rate due to an isotropic point source at arbitrary positions can be well approximated using the library with acceptable error. Thus, the library can be used for efficient analysis of radiation streaming due to arbitrary distributions of gamma-ray sources.

• The Journal of the Korea Contents Association
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• v.16 no.10
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• pp.758-765
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• 2016

This study analyzed for the radioactive shielding wall, which shields the medical linear accelerator. This allows to evaluate the level of waste with respect to the shield wall, which accounts for more than half of the cost of dismantling later linac facility. In addition, by analyzing the waste processing method according we discuss the way to obtain the benefits in terms of dismantling cost. Results of the simulate, the amount sufficient to screen the amount of neutron radiation occurring in the shielding wall linac was measured. And neutron activation analysis results were analyzed nuclides more than about 20. This analysis was in excess of that, $^{24}Na$, $^{45}Ca$, $^{59}Fe$ nucleus paper deregulation concentration. The value is reduced is greater the deeper the depth of the shielding wall concentration. Based on this, three specific areas (E, F, G) was estimated to be impossible to landfill or recycling. The rest area was estimated to be buried or recycled if possible more than a predetermined depth.

• 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.3 no.4
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• pp.185-197
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• 1971

Korea's TRIGA Mark-Ⅱ reactor was primarily designed in 1950's and was constructed in 1962 for 100 kw thermal output, but it was upgraded to 250 kw in July 1969. Nevertheless, the shield remains unchanged, although the radiation level has increased. The result of computation On this paper shows that, with the existing shield, it is safe for the fast neutrons even after the power upgrading by 2.5 times. It is, however, somewhat dangerous for the gamma rays which are comprised of primary and secondary. For the analysis of the reactor shielding design, an attempt is made for the computation toward the horizontal direction. From theoretical point of view, it can be concluded that some layer of additional shield must be reinforced to the existing concrete in order to be radiologically safe in the reactor hall.