• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.05a
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• pp.293-294
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• 2017

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.4
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• pp.511-521
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• 2022

Decommissioning of nuclear power plants generates a large amount of radioactive waste in a short period. Moreover, Radioactive waste has various forms including a large volumes of metal, concrete, and solid waste. The disposal of decommissioning waste using 200 L drums is inefficient in terms of economics, work efficiency, and radiation safety. Therefore, The Korea Radioactive Waste Agency is developing large containers for the packaging, transportation, and disposal of decommissioning waste. Assessing disposability considering the characteristics of the radioactive waste and facility, convenience of operation, and safety of workers is necessary. In this study, the exposure dose rate of workers during the disposal of new containers was evaluated using Monte Carlo N-Particle Transport code. Six normal and four abnormal scenarios were derived for the assessment of the dose rate in a near surface disposal facility operation. The results showed that the calculated dose rates in all normal scenarios were lower than the direct exposure dose limitation of workers in the safety analysis report. In abnormal scenarios, the work hours with dose rates below 20 mSv·y^-1 were calculated. The results of this study will be useful in establishing the optimal radiation work conditions.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.220-225
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• 2003

Generally, liquid radioactive wastes generated in nuclear power plant exist in powder form which do not contain moisture through the evaporating process of the Liquid Waste Management System and drying process of the Concentrated Waste Drying System. This powder form wastes are blended homogeneously with paraffin solidification agent and packed in metal drum to ensure its stability during handling and disposal. However, it was experienced that the powder form wastes were not blended homogeneously and separated into two layers in metal drum, on the other hand, a Portion of powder was adhered and solidified to the Inside parts of facility during the blending process. And the flaw of blending process above would come in case the mole ratio of Boron to Sodium in liquid radioactive wastes exceeds 0.2.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.1043-1050
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• 2020

Radioactive aerosols are produced during the cutting of contaminated and activated metals. They must be collected and removed by a high-performing filtration system before releasing to the environment from the decommissioning workplace. The filtration system requires regular replacement to ensure the sufficient removal of radioactive aerosols because its filtration efficiency gradually decreases. This study evaluates the efficiency and lifetime of filters while cutting metals by using a plasma arc cutter. Particularly, this study considers the aerodynamic diameter distribution of number and mass concentrations for aerosols from 6 nm to 10 ㎛ when evaluating the performance of filters. After 20 time reuses for cutting operation performed in a cutting chamber, the removal efficiency is reduced from over 99 to below 93% at 2 ㎛. The results are used to analyze the lifetime of filters, the frequencies of their replacements, and impact on internal radiation dose.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1805-1815
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• 2019

While nickel-based alloys have been widely used for power plants due to corrosion resistance and good mechanical properties, during the last couple of decades, failures of nuclear components increased gradually. One of main degradation mechanisms was primary water stress corrosion cracking at dissimilar metal welds of piping and reactor head penetrations. In this context, precise estimation of welding effects became an important issue for ensuring reliability of them. The present study deals with a series of finite element analyses and crack growth rate evaluation of Alloys 690/152. Firstly, variation of residual stresses and equivalent plastic strains was simulated taking into account welding of a cylindrical block. Subsequently, extraction and pre-cracking of compact tension (CT) specimens were considered from different locations of the block. Finally, crack growth curves of the alloys and heat affected zone were developed based on analyses results combined with experimental data in references. Characteristics of crack growth behaviors were also discussed in relation to mechanical and fracture parameters.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.278-284
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• 2004

The processes for the smelting of a metal product and the solidification of a molten salt were developed respectively to treat the products from the electrochemical reduction process. The method for the separation of a metal product in a magnesia container from the residual. salt and consequent smelting of it to a metal ingot by the multi step heating in vacuum was proposed. The new concept using a dual vessel and a salt valve was also suggested for the solidification of a molten salt into a regular size and shape which is suitable for the transport and measurement. The results obtained in the study will be applied to the design of the hot cell demonstration system of the Advanced Spent Fuel Conditioning Process of KAERI.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.54-59
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• 2003

Several radionuclides are considered as an object of the assesment to develop a scaling factor and a periodical verification method which are needed for the evaluation of radionuclide inventory of various radioactive wastes from nuclear power plants in Korea. A selective separation of $^{55}Fe$, $^{90}FSr$ 및 $^{94}Nb$ which should be recovered individually for the radiochemical analysis was described in detail. Sorption and desorption behaviours of ion exchange and extraction chromatographic resins for Fe, Sr, Nb and co-existing metal ions were Investigated using an artificial waste solution simulated of chemical composition of real radioactive wastes. Separation conditions available for the sequential recovery of these metal ions from a single sample were optimized to minimize a discharge of radioactive wastes produced through the analytical process and a radiation exposure to analysts. Their recovery yields were measured with reliability.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.411-422
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• 2016

Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R&D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0~4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.

• Resources Recycling
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• v.6 no.4
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• pp.38-47
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• 1997

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.462.2-462.2
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• 2003