• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.655-659
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• 2003

The research reactor (TRIGA Mark-III(KRR-2)) was constructed and had been operated in 1972. In 1999 the radioisotope process units had stopped its operation due to normal operation of HANARO. In 2003 the core assembly was decommissioned by D&D program. The contact exposure rate on the core assembly and the rotary specimen rack are from 300mSv/h to 700mSv/h. This report describes the decontaminationing procedures, the health physics programs, and the waste management.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2007.05a
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• pp.190-191
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• 2007

Cleanout of the F Reactor Fuel Storage Basin (FSB) is an element of the FSB decontamination and decommissioning (D&D) and is required to complete interim safe storage (ISS) of the F Reactor. Following reactor shutdown and in preparation for a deactivation layaway action in 1970, the water level in the FReactor FSB was reduced to approximately 0.6 m (2 ft) over t]to floor. Basin components and other miscellaneous items were left or placed in the FSB. The item placement was performed with a sense of finality, and no attempt was made to place the items in an orderly manner. The F Reactor FSB was then filled to grade level with 6(20of local surface material (essentially a fine sand). The reactor FSB backfill cleanout has the potential of having to remove spent nuclear fuel (SNF) that may have been left unintentionally. Based on previous cleanout of six water-filled FSBs with similar designs (i.e., the B, C, D, and DR FSBs in the 1980's), it was estimated that up to five SNF elements could be discovered in the F FSB (I). In reality about 17 full SNF elements were found in the excavation. This paper covers the technical and programmatic challenges of performing this decommissioning effort with some of the controls used for SNF management. The paper also will highlight how many various technologies were married into a complete package to address the issue at hand and show how no one tools could complete the job, but combined, good progress is being made.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.203-212
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• 2019

Disposal nonconformity of radioactive wastes refers to radioactive wastes that need to be treated, solidified and packaged during operation or decommissioning of NPPs, and are typically exemplified by particulate radioactive wastes with dispersion characteristics. These wastes include the dried powders of concentrated wastes generated in the process of operating NPPs, slurry and sludge, various powdered wastes generated in the decommissioning process (crushed concrete, decontamination sludge, etc.), and fine radioactive soil, which is not easy to decontaminate. As these particulate wastes must be packaged so that they become non-dispersive, they are solidified with solidification agents such as cement and polymer. If they are treated using existing solidification methods, however, the volume of the final wastes will increase. This drawback may increase the disposal cost and reduce the acceptability of disposal sites. Accordingly, to solve these problems, this study investigates the pelletization of particulate radioactive wastes in order to reduce final waste volume.

• Journal of Radiation Protection and Research
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• v.48 no.1
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• pp.44-51
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• 2023

Background: Aptamers are currently being used in various fields including medical treatments due to their characteristics of selectively binding to specific molecules. Due to their special characteristics, the aptamers are expected to be used to remove radionuclides from a large amount of liquid radioactive waste generated during the decommissioning of nuclear power plants. The radiological effects on the aptamers should be evaluated to ensure their integrity for the application of a radionuclide removal technique. Materials and Methods: In this study, Monte Carlo N-Particle transport code version 6 (MCNP6) and Monte Carlo damage simulation (MCDS) codes were employed to evaluate the radiological effects on the aptamers. MCNP6 was used to evaluate the secondary electron spectrum and the absorbed dose in a medium. MCDS was used to calculate the DNA damage by using the secondary electron spectrum and the absorbed dose. Binding experiments were conducted to indirectly verify the results derived by MCNP6 and MCDS calculations. Results and Discussion: Damage yields of about 5.00×10^-4 were calculated for 100 bp aptamer due to the radiation dose of 1 Gy. In experiments with radioactive materials, the results that the removal rate of the radioactive ⁶⁰Co by the aptamer is the same with the non-radioactive ⁵⁹Co prove the accuracy of the previous DNA damage calculation. Conclusion: The evaluation results suggest that only very small fraction of significant number of the aptamers will be damaged by the radioactive materials in the liquid radioactive waste.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.377-387
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• 2018

Korea's first commercial nuclear power plant, Kori Unit 1, was permanently shut down on June 18, 2017, after 40 years of successful operation. Kori Unit 1 plans to construct a waste treatment facility in the turbine building prior to commencement of dismantling in earnest. Various radioactive wastes are decontaminated, disassembled, cut and melted in the waste treatment facility and sent to the radioactive waste repository. The proportion of metal radioactive waste in dismantled waste is about 70%, of which large metal radioactive waste is mainly generated in the primary circuit and has high radioactivity, so radiation exposure must be managed during disassembly. In this study, the steam generators are selected as large metal radioactive waste, the exposure doses of the dismantling workers are calculated using RESRAD-RECYCLE code and the methods for reducing the exposure doses are suggested.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.671-675
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• 2003

In accordance with the KRR-1 & 2 decommissioning project, the decontamination and dismantling activities of the KRR-2 auxiliary facilities, radioisotope production facilities, were completed from Aug 2001 to Dec 2002. The auxiliary facilities were composed of the concrete hot-cell, lead hot-cells and several laboratories for the radioisotope production. The dismantling objects are home hoods, experimental desks, sinks, and contaminated inner facilities. For the purpose of the safe decommissioning activity, the method statements and working procedures were set up. The manpower of the total 20,933 man-hour was required and several dismantling equipments were also. The maximum surface contamination is: 9.24 Bq/$\textrm{cm}^2$ in removable contamination and 350,000 cpm in fixed contamination. The total amount of 62.146 Ton was raised as dismantled waste with kinds of the concretes, wood, steels, etc. The collective dose was evaluated as 0.33 mam-mSv during this period.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.556-561
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• 2017

Serving as an excellent adsorbent and inorganic ion exchanger in the water purification field, zeolite 4A has in this work presented a strong capability for purifying radioactive waste, such as $Sr^{2+}$, $Cs^+$, and $Co^{2+}$ in water. During the processes of decontamination and decommissioning of suspended solids and organics in low-level radioactive wastewater, the purification performance of zeolite 4A has been studied. Under ambient temperature and neutral condition, zeolite 4A absorbed simulated radionuclides such as $Sr^{2+}$, $Cs^+$, and $Co^{2+}$ with an absorption rate of almost 90%. Additionally, in alkaline condition, the adsorption percentage even approached 98.7%. After conducting research on suspended solids and organics of zeolite 4A for the treatment of radionuclides, it was found that the suspended clay was conducive to absorption, whereas the absorption of organics in solution was determined by the species of radionuclides and organics. Therefore, zeolite 4A has considerable potential in the treatment of radioactive wastewater.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.3
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• pp.291-301
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• 2015

Product regulations specify requirements or constraints for products that manufacturers must comply with across the entire product lifecycle, from design and manufacture, through operation and maintenance, to recycling and disposal. This paper suggests a system framework and its essential components for the verification of a pressure vessel design using the industrial standards of Korea. The authors also present methods to generate design template data from legacy design systems and to construct a regulation knowledge base. The proposed framework is demonstrated through experiments involving pressure vessel design verification using a prototype system.

• IEIE Transactions on Smart Processing and Computing
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• v.6 no.1
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• pp.66-70
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• 2017

Radioactive materials are used in medicine, non-destructive testing, and nuclear plants. Source localization is especially important during nuclear decommissioning and decontamination because the actual location of the radioactive source within nuclear waste is often unknown. The coded-aperture imaging technique started with space exploration and moved into X-ray and gamma ray imaging, which have imaging process characteristics similar to each other. In this study, we simulated $21{\times}21$ and $37{\times}37$ coded aperture collimators based on a modified uniformly redundant array (MURA) pattern to make a gamma imaging system that can localize a gamma-ray source. We designed a $21{\times}21$ coded aperture collimator that matches our gamma imaging detector and did feasibility experiments with the coded aperture imaging system. We evaluated the performance of each collimator, from 2 mm to 10 mm thicknesses (at 2 mm intervals) using root mean square error (RMSE) and sensitivity in a simulation. In experimental results, the full width half maximum (FWHM) of the point source was $5.09^{\circ}$ at the center and $4.82^{\circ}$ at the location of the source was $9^{\circ}$. We will continue to improve the decoding algorithm and optimize the collimator for high-energy gamma rays emitted from a nuclear power plant.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.157-168
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• 2016

Radioactive waste container designs should comply with the requirements for safety (i.e., transportation, storage, disposal) and other criteria such as economics and technology. These criteria are also applicable to the future management of the large amount of LLW and VLLW to arise from decontamination and decommissioning (D&D) of nuclear power plants, which have different features compared to that of wastes from operation and maintenance (O&M). This paper proposes to develop a set of standard containers of multi-purpose usage for transportation, storage and disposal. The concepts of the containers were optimized for management of D&D wastes in consideration of national system for radioactive waste management, in particular the Gyeongju Repository and associated infrastructures. A set of prototype containers were designed and built : a soft bag for VLLW, two metallic containers for VLLW/LLW (a standard IP2 container for sea transport and ISO container for road transport). Safety analyses by simulation and tests of these designs show they are in compliance with the regulatory requirements. A further development of a container with concrete is foreseen for 2016.