• Journal of the Korean Society of Radiology
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• v.11 no.4
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• pp.189-195
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• 2017

Radiation protection equipment has widely used to protect human body from radiations, for example X-ray and gamma ray. The material of the radiation protection equipment is mainly lead (Pb) which has brought out lead poisoning and pollution when the equipment is fallen into disuse. This problem makes research and development find new Pb-free materials for use of radiation protection. Manufacturing and evaluation processes for developing those material were carried out repletely until obtaining the performance of protection rate. In this study, combination possibility of shielding material was studied using Geant4 monte carlo simulation. X-ray tube under the same condition in the real measurement of the protection rate was simulated, and X-ray tube spectrum was obtained. The X-ray tube spectrum was applied to study on the protection rate and lead equivalent. The porosity effect was simulated, and was one of key factors to determine protection rate or lead equivalent in radiation protection sheet of Pb-free.

• 대한방사선방어학회:학술대회논문집
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• 2010.04a
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• pp.70-71
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• 2010

• Journal of Radiation Protection and Research
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• v.41 no.2
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• pp.117-121
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• 2016

Background: Definition and grouping of initiating events (IEs) are important basics for probabilistic safety assessment (PSA). An IE in a spent fuel reprocessing plant (SFRP) is an event that probably leads to the release of dangerous material to jeopardize workers, public and environment. The main difference between SFRPs and nuclear power plants (NPPs) is that hazard materials spread diffusely in a SFRP and radioactive material is just one kind of hazard material. Materials and Methods: Since the research on IEs for NPPs is in-depth around the world, there are several general methods to identify IEs: reference of lists in existence, review of experience feedback, qualitative analysis method, and deductive analysis method. While failure mode and effect analysis (FMEA) is an important qualitative analysis method, master logic diagram (MLD) method is the deductive analysis method. IE identification in SFRPs should be consulted with the experience of NPPs, however the differences between SFRPs and NPPs should be considered seriously. Results and Discussion: The plutonium uranium reduction extraction (Purex) process is adopted by the technics in a model reprocessing plant. The first extraction cycle (FEC) is the pivotal process in the Purex process. Whether the FEC can function safely and steadily would directly influence the production process of the whole plant-production quality. Important facilities of the FEC are installed in the equipment cells (ECs). In this work, IEs in the FEC process were identified and categorized by FMEA and MLD two methods, based on the fact that ECs are containments in the plant. Conclusion: The results show that only two ECs in the FEC do not need to be concerned particularly with safety problems, and criticality, fire and red oil explosion are IEs which should be emphatically analyzed. The results are accordant with the references.

• 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.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.424-435
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• 2016

Background: The International Commission on Radiological Protection (ICRP) recommendations and the Federal Guidance Report (FGR) published by the U.S. Environmental Protection Agency (EPA) have been widely applied worldwide in the fields of radiation protection and dose assessment. The dose conversion coefficients of the ICRP and FGR are widely used for assessing exposure doses. However, before the coefficients are used, the user must thoroughly understand the derivation process of the coefficients to ensure that they are used appropriately in the evaluation. Materials and Methods: The ICRP provides recommendations to regulatory and advisory agencies, mainly in the form of guidance on the fundamental principles on which appropriate radiological protection can be based. The FGR provides federal and state agencies with technical information to assist their implementation of radiation protection programs for the U.S. population. The system of radiation dose assessment and dose conversion coefficients in the ICRP and FGR is reviewed in this study. Results and Discussion: A thorough understanding of their background is essential for the proper use of dose conversion coefficients. The FGR dose assessment system was strongly influenced by the ICRP and the U.S. National Council on Radiation Protection and Measurements (NCRP), and is hence consistent with those recommendations. Moreover, the ICRP and FGR both used the scientific data reported by Biological Effects of Ionizing Radiation (BEIR) and United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) as their primary source of information. The difference between the ICRP and FGR lies in the fact that the ICRP utilized information regarding a population of diverse races, whereas the FGR utilized data on the American population, as its goal was to provide guidelines for radiological protection in the US. Conclusion: The contents of this study are expected to be utilized as basic research material in the areas of radiation protection and dose assessment.

• Journal of Microbiology and Biotechnology
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• v.28 no.4
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• pp.527-533
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• 2018

This study compared the radioprotective effects of high-molecular-weight poly-gamma-glutamate (${\gamma}-PGA$, average molecular mass 3,000 kDa) and a reduced form of glutathione (GSH, a known radioprotector) on calf thymus DNA damage. The radiation-induced DNA damage was measured on the basis of the decreased fluorescence intensity after binding the DNA with ethidium bromide. All the experiments used $^{60}Co$ gamma radiation at 1,252 Gy, representing 50% DNA damage. When increasing the concentration of ${\gamma}-PGA$ from 0.33 to $1.65{\mu}M$, the DNA protection from radiation-induced damage also increased, with a maximum of 87% protection. Meanwhile, the maximal DNA protection when increasing the concentration of GSH was only 70%. Therefore, ${\gamma}-PGA$ exhibited significant radioprotective effects against gamma irradiation.

• Journal of Radiation Protection and Research
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• v.34 no.3
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• pp.144-150
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• 2009

This study examines how much the radiation dose rate around it varies if a crack occurs on the spent nuclear fuel rod. The spent nuclear fuel rod to be examined is that of Kori unit 3&4. The source terms are evaluated using the ORIGEN-ARP that is part of the version 5.1 of the SCALE package. The radiation dose rate is assessed using the TORT. To check if the structure of a fuel rod is appropriately modeled in the TORT calculation, the calculation results by the TORT are compared with those by the ANISN for the same case. From the code simulation, it is known that if a crack occurs on the spent nuclear fuel rod, the neutron dose rate varies depending on what material is the crack filled with, but the gamma dose rate varies irrespective of type of the material that the crack is filled with.

• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.310-314
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• 2016

Background: FCTC10 container is designed to transport $^{60}Co$ radioactive sources used in irradiation industry. It belongs to Type B(U) Category III (yellow) package when being loaded with a $^{60}Co$ source of $1.8{\times}10^5$ Ci. Materials and Methods: The container is constituted of shielding container, basket, protective cover and bracket. Shielding ability is provided mainly by stainless steel shells, tungsten alloy and lead among steel shells. Radiation level around the container has been calculated with both Monte Carlo simulations and measurements. Results and Discussion: It is proven that the shielding performance of the container fulfills the requirements in GB11806-2004 (Regulations for the safe transport of radioactive material, China Standard Press). Exposure doses to workers and to critical groups of public were calculated based on hypothetical exposure scene according to transport practice experience. Conclusion: The results show that doses to workers and public are less than the constraint dose considered in design, and the radiation level would be increased less than a factor of 2 under design basis accidents.

• Advanced Industrial SCIence
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• v.2 no.3
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• pp.1-7
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• 2023

Recently, it is increasing that a issue of concern about radiation exposure. It affects soil, water, air, crops, etc., and in the long term, environmental pollution and food pollution occur, and it is considered to cause social problems and economic damage. Radiation exposure causes diseases and health problems, but as a method for diagnosing diseases, nuclear medicine tests such as X-ray imaging, CT, and PET-CT are conducted, and radiation isotopes are exposed for the purpose of cancer treatment. A Hungarian case study on radiation in water, particularly drinking water, following the release of radioactive waste from Fukushima, and an examination of the Larsemann Hills area in Antarctica, found that it was within the prescribed radioactivity limits of drinking water recommended by the World Health Organization. We looked at radioprotective agents, focusing on DNA damage, cell and organ damage, and cancer, and also investigated various literatures on ACE inhibitors, antioxidants, and natural substances among restoration materials. Although exposed to radiation in everyday life, the reason why it can be safe is probably because there is a radiation protection material and a recovery material for radiation exposure, so we are trying to find possible materials.

• Journal of the Korean Society of Radiology
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• v.15 no.2
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• pp.191-199
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• 2021

The purpose of this research is to analyze the radiation shielding effect of soft magnetic material to confirm the applicability to the military facilities. The soft magnetic material is known to be effective in shielding EMP. If this material is also effective in radiation shielding, it is expected that it has a lot of applicability in military protection. In particular, this material contains boron, so it will be effective in shielding neutrons. In this research, experiments were conducted using Cs-137 and Co-60 sources to check the gamma ray shielding effect. In addition, the Monte Carlo N-Particle(MCNP) modeling was applied to evaluate the gamma ray and neutron shielding effect of a military command tent. As a result, as the soft magnetic thickness increased, the shielding performance improved according the linear attenuation law of gamma ray and neutron. Therefore, this research verified that the application of soft magnetic material for military purposes in radiation shielding would be effective.