• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.107-120
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• 2019

The decommissioning of one nuclear power plant in a multi-unit nuclear power plant (multi-unit NPP) site may pose radiation exposure risk to decommissioning workers. Thus, it is essentially required to evaluate the exposure dose of decommissioning workers of operating multi-unit NPPs nearby. The ENDOS program is a dose evaluation code developed by the Korea Atomic Energy Research Institute (KAERI). As two sub-programs of ENDOS, ENDOS-ATM to anticipate atmospheric transport and ENDOS-G to calculate exposure dose by gaseous radioactive effluents are used in this study. As a result, the annual maximum individual dose for decommissioning workers is estimated to be $2.31{\times}10^{-3}mSv{\cdot}y^{-1}$, which is insignificant compared with the effective dose limit of $1mSv{\cdot}y^{-1}$ for the public. Although it is revealed that the exposure dose of operating multi-unit NPPs does not result in a significant impact on decommissioning workers, closer examination of the effect of additional exposure due to actual demolition work is required. The calculation method of this study is expected to be utilized in the future for planned decommissioning projects in Korea. Because domestic NPPs are located in multi-unit sites, similar situations may occur.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.1-13
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• 2019

The accident at the Fukushima Dai-ichi Nuclear Power Plant (FDNPP) resulted in the deposition of large quantities of radionuclides over parts of eastern Japan. Radioactive contaminants have been observed over a large area including forests, cities, rivers and lakes. Due to the strong adsorption of radioactive cesium by soil particles, radioactive cesium migrates with the eroded soil, follows the surface flow paths, and is delivered downstream of population-rich regions and eventually to coastal areas. In this study, we developed a model to simulate the transport of contaminated sediment in a watershed hydrological system and this model was compared with observation data from eroded soil observation instruments located at the Korea Atomic Energy Research Institute. Two methods were applied to analyze the soil particle size distribution of the collected soil samples, including standardized sieve analysis and image analysis methods. Numerical models were developed to simulate the movement of soil along with actual rainfall considering initial saturation, rainfall infiltration, multilayer and rain splash. In the 2019 study, a numerical model will be used to add rainfall shield effect by trees, evaporation effect and shield effects of surface water. An eroded soil observation instrument has been installed near the Wolsong nuclear power plant since 2018 and observation data are being continuously collected. Based on these observations data, we will develop the numerical model to analyze long-term behavior of radionuclides on land as they move from land to rivers, lakes and coastal areas.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.59-73
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• 2019

A decontamination technology for radioactive liquid wastes was newly developed and hypothetically applied to the liquid waste management system (LWMS) of the nuclear power plant (NPP) to evaluate its decontamination efficacy for the purpose of the fundamental reduction of spent resins. The basic principle of the developed technology is to convert major radionuclide ions in the liquid wastes into inorganic crystal minerals via chemical or biological techniques. In a laboratory batch experiment, the biological method selectively removed more than 80% of cesium within 24 hours, and the chemical method removed more than 95% of cesium. Other major nuclides (Co, Ni, Fe, Cr, Mn, Eu), which are commonly present in nuclear radioactive liquid wastes, were effectively scavenged by more than 99%. We have designed a module including the new technology that could be hypothetically installed between the reverse osmosis (R/O) package and the organic ion-exchange resin in the LWMS of the APR1400 reactor. From a technical evaluation for the virtual installation, we found that more than 90% of major radionuclides in the radioactive liquid wastes were selectively removed, resulting in a large volume reduction of spent resins. This means that if the new technology is commercialized in the future, it could possibly provide drastic cost reduction and significant extension of the life of resins in the management of spent resins, consequently leading to delay the saturation time of the Wolsong repository.

• Journal of Radiation Protection and Research
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• v.47 no.1
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• pp.1-7
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• 2022

Background: In Japan, new regulations that revise the dose limit for the lens of the eye (hereafter the lens), operational quantities, and measurement positions for the lens dose were enforced in April 2021. Based on the international safety standards, national guidelines, the results of the Radiation Safety Research Promotion Fund of the Nuclear Regulation Authority, and other studies, the Working Group of Radiation Protection Standardization Committee, the Japan Health Physics Society (JHPS) developed a guideline for radiation dose monitoring for the lens. Materials and Methods: The Working Group of the JHPS discussed the criteria of non-uniform exposure and the management criteria set not to exceed the dose limit for the lens. Results and Discussion: In July 2020, the JHPS guideline was published. The guideline consists of three parts: main text, explanations, and 26 examples. In the questions, the corresponding answers were prepared, and specific examples were provided to enable similar cases to be addressed. Conclusion: With the development of the guideline on radiation dose monitoring of the lens, radiation managers and workers will be able to smoothly comply with revised regulations and optimize radiation protection.

• Journal of Korea Water Resources Association
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• v.55 no.10
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• pp.785-799
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• 2022

The aim of this study was to characterize the emission pathways and the footprint of greenhouse gases (GHG) in Daecheong Reservoir using the G-res Tool, and to evaluate the GHG emission intensity (EI) compared to other energy sources. In addition, the change in GHG emissions was assessed in response to the total phosphorus (TP) concentration. The GHG flux in post-impoundment was found to be 262 gCO₂eq/m²/yr, of which CO₂ and CH₄ were 45.7% and 54.2%, respectively. Diffusion of CO₂ contributed the most, followed by diffusion, degassing, and bubbling of CH₄. The net GHG flux increased to 510 gCO₂eq/m²/yr because the forest (as CO₂ sink) was lost after dam construction. The EI of Daecheong Reservoir was 86.8 gCO₂eq/kWh, which is 3.7 times higher than the global EI of hydroelectric power, due to its low power density. However, it was remarkable to highlight the value to be 9.5 times less than that of coal, a fossil fuel. We also found that a decrease in TP concentration in the reservoir leads to a decrease in GHG emissions. The results can be used to improve understanding of the GHG emission characteristics and to reduce uncertainty of the national GHG inventory of dam reservoirs.

• Journal of the Korean Society of Radiology
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• v.18 no.5
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• pp.491-498
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• 2024

Preclinical positron emission tomography (PET) requires excellent spatial resolution because the subject of imaging is a very small animal. To achieve this, a detector is configured using fine scintillation pixels. In this study, we aim to increase the scintillation pixel array by processing the bottom surface of the scintillation pixels differently from the array of scintillation pixels that can be imaged in the same photosensor performed in the previous study. To this end, we designed a detector using DETECT2000, which can simulate light in the scintillator, and performed a simulation. The detector was configured from an 11 × 11 array to a 16 × 16 array, and the bottom surface was configured as a polished surface (POLISH) and a rough surface (GROUND) to obtain a flood image. As a result, it was confirmed that the scintillation pixel images were better separated on the GROUND surface than on the POLISH surface as the scintillation pixel array expanded. Furthermore, on the GROUND surface, it was confirmed that the peaks of the scintillation pixel images in the corner area were separated and imaged even in the 16 × 16 array.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.2
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• pp.77-84
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• 2005

A voloxidizer for a hot cell demonstration, that handles spent fuels of a high radiation level in a limited space should be small and spent fuel powders should not be dispersed out of the equipment involved. In this study a density rate equation as well as the Stokes'equation has been proposed in order to obtain the theoretical terminal velocity of powders. The terminal velocity of U$_{3}$O$_{8}$ has been predicted by using the terminal velocity of SiO$_{2}$, and then determination has been the optimum air flow rate which is able to prevent powders from scattering. An equation which has shown a relationship between theoretical terminal velocities of U$_{3}$O$_{8}$ and SiO$_{2}$ has been derived with the help of the Stokes'equation, and then an experimental verification made for the theoretical Stokes' equation of SiO$_{2}$ by means of an experimental device made of acryl. The theoretical terminal velocity based on the proposed density rate equation has been verified by detecting U$_{3}$O$_{8}$ powders in a filter installed in the mock-up voloxidizer. As the results, the optimum air flow rates seem to be 20 LPM by the Stokes'equation while they are 14.5 L/min by the density rate equation. At the experiments with the mock-up voloxidizer, a trace amount of U$_{3}$O$_{8}$ seems to be detectable at the air flow rate of 14.5 L/min by the density rate equation, but U$_{3}$O$_{8}$ powders of 7$\mu$m diameter seem detectable at the air flow rate of 20 L/min by the Stokes'equation. It is revealed that 14.5 L/min is the optimum air flowe rate which is capable of preventing U$_{3}$O$_{8}$ powders from scattering in the UO$_{2}$ voloxidizer and the proposed density rate equation is proper to calculate the terminal velocity of U$_{3}$O$_{8}$ powders.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.3
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• pp.131-139
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• 2011

This study was carried out to examine the stability of $Na_2CO_3-H_2O_2$ carbonate solution with aging time in the dissolving solution after oxidative dissolution of U by a carbonate solution, the Cs/Re filtrate after selective precipitation of Cs and Re (as a surrogate for Tc), and the acidification filtrate after precipitation of U by acidification, respectively. The compositions of dissolving solution were not changed with ageing time, and the selective precipitation of Re and Cs was also not affected without regard to the aging time of dissolving solution. The successive removal of Cs and Re from a carbonate dissolving solution was possible. However, the recovery yield of U by acidification was decreased with increasing the aging time of the dissolving solution and the Cs/Re-filtrate, respectively, because U was converted from the uranyl peroxocarbonato complex to the uranyltricarbonate in the solution aged for a long time. Accordingly, it is effective that a dissolving solution and a Cs/Re filtrate are treated within the aging of 7 days, respectively, in order to recover U more than 99%. On the other hand, the recovery of U was carried out in order of the oxidative dissolution of U selective precipitation of Re and Cs precipitation of U by acidification. Almost all of U and a part of FP were co-dissolved in oxidative dissolution step. Over 99% of Re and Cs from the FP co-dissolved with U could be removed by a TPPCl (tetraphenylphosphonium chloride) and a NaTPB (sodium tetraphenylborate), respectively. U was precipitated nearly 100% by acidification to pH 4. Therefore, it was confirmed that the validity of recovery of U by precipitation methods from a SF (spent fuel) in the $Na_2CO_3-H_2O_2$ solution.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.171-177
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• 2007

This study suggested a new method to stabilize molten salt wastes generated from the pyre-process for the spent fuel treatment. Using conventional sol-gel process, $SiO_2-Al_2O_3-P_2O_5$ (SAP) inorganic material that is reactive to metal chlorides were prepared. In this paper, the reactivity of SAP with the metal chlorides at $650{\sim}850$, the thermal stability of reaction products and their leach-resistance under the PCT-A test method were investigated. Alkali metal chlorides were converted into metal aluminosilicate($LixAlxSi1-_xO_{2-x}$) and metal phosphate($Li_3PO_4\;and\;Cs_2AlP_3O_{10}$) While alkali earth and rare earth chlorides were changed into only metal phosphates ($Sr_5(PO_4)_3Cl\;and\;CePO_4$). The conversion rate was about $96{\sim}99%$ at a salt waste/SAP weight ratio of 0.5 and a weight loss up to $1100^{\circ}C$ measured by thermogravimetric analysis were below 1wt%. The leach rates of Cs and Sr under the PCT-A test condition were about $10^{-2}g/m^2\;day\;and\;10^{-4}g/m^2\;day$. From these results, it could be concluded that SAP can be considered as an effective stabilizer for metal chlorides and the method using SAP will give a chance to reduce the volume of salt wasteform for the final disposal through further researches.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.255-263
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• 2006

Vol-oxidizer is a device to convert $UO_2$ pellets into $U_3O_8$ powder and to feed a homogeneous powder into a Metal Conversion Reactor in the ACP(Advanced Spent Fuel Conditioning Process). In this paper, we propose a design model of the vol-oxidizer, develop the new vol-oxidizer with a capacity of 20 kg HM/batch in $UO_2$ pellets, and conduct a verification for the device. Design considerations include the internal structure, the capacity, the heating position of the device, and the size. The dimensions of the new vol-oxidizer are decided by the design model. We determine a permeability test of the $U_3O_8$ measuring the temperature distribution, and the volume of $UO_2$ and $U_3O_8$. We manufactured the new vol-oxidizer for a 20 kg HM/batch in $UO_2$ pellets, and then analyzed the characteristics of the $U_3O_8$ powder for the verification. The experimental results show that the permeability of the $U_3O_8$ throughout mesh enhance more than old vol-oxidizer, the oxidation time takes only 8 hours when compared with the 13 hours of the old device, and the average distribution of particle size is $40{\mu}m$. The capacities of new vol-oxidizer for a 20 kg HM/batch in $UO_2$ pellets were agree well with the predictions of design model.