• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.79-86
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• 2009

Concrete materials in nuclear facilities may become contaminated or activated by various radionuclides through different mechanism. Decommissioning and dismantling of these facilities produce considerable quantities such as concrete structure, rubble. In this paper, the characteristics distribution of the radionuclide have been investigated for the effects of the heating and grinding test for aggregate size such as gravel, sand and paste from decommissioning of the TRIGA MARK II research reactor and uranium conversion plant. The experimental results showed that most of the radionuclide could be removed from the gravel, sand aggregate and concentrated into a paste. Especially, we found that the heating temperature played an important role in separating the radionuclide from the concrete waste. Contamination of concrete is mainly concentrated in the porous paste and not in the dense aggregate such as the gravel and sand. The volume reduction rate could be achieved about 80% of activated concrete waste and about 75% of dismantled concrete waste generated from UCP.

• 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 Nuclear Fuel Cycle and Waste Technology
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• v.1 no.1
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• pp.37-47
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• 2013

This paper describes basic plans for the development of a radioactive waste disposal facility with the introduction of Nuclear Power Plants (NPPs) for Kenya. The specific objective of this study was to estimate the total projected waste volumes of low- and intermediate-level radioactive waste (LILW) expected to be generated from the Kenyan nuclear power programme. The facility is expected to accommodate LILW to be generated from operation and decommissioning of nuclear power plants for a period of 50 years. An on-site storage capacity of 700 $m^3$ at nuclear power plant sites and a final disposal repository facility of more than 7,000 $m^3$ capacity were derived by considering Korean nuclear power programme radioactive waste generation data, including Kori, Hanbit, and APR 1400 nuclear reactor data. The repository program is best suited to be introduced roughly 10 years after reactor operation. This study is important as an initial implementation of a national LILW disposal program for Kenya and other newcomer countries interested in nuclear power technology.

• Journal of the Korean Recycled Construction Resources Institute
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• v.5 no.4
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• pp.427-434
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• 2017

The generation of construction waste can be divided into a decommissioning phase and a new construction phase, and most of the waste is generated at the decommissioning stage. However, recently, domestic new construction construction has expanded to 150 trillion yards per year, so construction work is increasing rapidly. Especially, as the size of the construction work with much waste of construction waste exceeds 100 trillion, the management of the amount of construction waste in the new construction site is required. Unlike the dismantling work site, the new construction site can separate waste generated by each property, and relatively low foreign matter content is generated. The purpose of this study was to investigate the amount of construction waste generated by new construction sites and to calculate the unit amount of construction waste based on this. In addition, since the existing unit cost is centered on concrete and mixed waste, we set the basic unit by setting synthetic resin, waste wood, and waste board as additional items. The basic unit survey was carried out to investigate the wastes according to the characteristics of each construction period. As a result of the survey, the new construction site showed that most wastes were discharged in the first 30% and after 70% of the process, and the ratio of mixed construction waste was as high as 45%. As a result of this study, it was found that about twice as much waste was produced as compared with the conventional standard product.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.2
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• pp.163-169
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• 2015

In many nuclear power plant sites in Korea, high density storage racks were installed in the spent fuel pool to expand the spent fuel storage capacity. Nevertheless, the capability of the Hanbit nuclear site will be saturated by 2024. Also, 10 NPPs will reach their design life expiration date by 2029. In the case of the US, SFPI (Spent Fuel Pool Island) operated temporarily as a spent fuel storage option before spent nuclear fuels were transported to an interim storage facility or a final disposal facility. As a spent fuel storage option after shutdown during decommissioning, the SFPI concept can be expected to have the following effects: reduced occupational exposure, lower cost of operation, strengthened safety, and so on. This paper presents a case study associated with the regulations, operating experiences, and systems of SFPI in the US. In conclusion, the following steps are recommended for applying SFPI during decommissioning in Korea: confirmation of design change scope of SFPI and expected final cost, the submission of a decommissioning plan which is reflected in SFPI improvement plans, safety assessment using PSR, application of an operating license change for design change, regulatory body review and approval, design change, inspection by the regulatory body, education and commissioning for SFPI, SFPI operation and periodic inspection, and dismantling of SFPI.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.3010-3016
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• 2023

In recent years, the development of radioactive waste containers for nuclear facility decommissioning and dismantling is a critical issue because the Taiwan domestic boiling water reactor nuclear power plant is going to be decommissioned. The main purpose of this research is to design a metal container that meets the structural requirements of related regulations. At first, the shielding analysis was performed by varying dimensions of radioactive waste to determine the storage efficiency of the container. Then, a series of structural analyses for operational and accidental conditions of the container with full load were conducted, such as lifting, stacking, and drop impact conditions. On the other hand, the field drop impact tests were carried out to ensure structural integrity. The present research demonstrates the structural safety of the developed container for decommissioned nuclear facilities in Taiwan.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2489-2497
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• 2023

Co and Eu impurities in the SSCs are nuclides that dominantly influence the neutron-induced radioactive inventory in metal and concrete radwastes (radioactive wastes) during NPP decommission. The impurity concentrations provided by NUREG/CR-3474 were used for the practical range of Co and Eu impurity concentrations to be applied to the code calculations. Metal structures near the core were evaluated to be ILW (intermediate-level waste) for the whole range of Co impurity concentration, so the boundary line between ILW and LLW (low-level waste) has no change for the whole concentration range provided by NUREG/CR-3474. Also, the boundary line between VLLW (very low-level waste) and CW (clearance waste) in the concrete shield could alter a little depending on the Eu impurity concentration within the range provided by NUREG/CR-3474. From this work, it is found that the concentration of material impurities of SSCs gives no critical impact on determining radwaste levels.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.69-75
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• 2004

DORT and ORIGEN2 code were used for calculation of neutron flux and inventory in reactor pressure vessel(RPV) of Kori unit-1, To calculate neutron flux using DORT code, the reactor was divided into 94 mesh from the center of core to RPV and from 0 to 45 degree along the azimuth. The cross-sections of main nuclides were recalculated using neutron flux in the RPV region. The results showed that 95% of the total activity in RPV came from the nuclides of $^{55}$ Fe, $^{60}$ Co, $^{59}$ Ni and $^{63}$ Ni. And the total activity with cooling of more than 50 years after decommissioning was no more than 0.2% of at the time of shutdown. Considering the weight of RPV is 210 tons, the initial total activity of RPV reached 5.25${\times}$10$^{6}$ GBq. To verify results of ORIGEN2 calculation, comparison between calculated and measured value at RPV of Kori unit-1 was peformed. The comparison results showed a good agreement.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.2
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• pp.87-96
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• 2012

With a view to determine a safe speed the limit of a manipulator arm, several experiments was performed with a multi-jointed manipulator in maintenance and decommissioning tasks of nuclear facilities. Under the simulated emergency conditions, which were generated with random combinations of manipulator arm speed, failure probability and failure type, response characteristics of human operators to various malfunctions of a manipulator arm were measured in terms of reaction time, number of false alarm, and number of misses. This paper demonstrated that failure type, manipulator axes and manipulator arm speed has significant effects on human reaction time. As a whole the reaction time was slightly increased with manipulator arm speed, which is showed somewhat different pattern due to failure type. The reaction time to an axis acting on a workpiece directly, which could flex and extend, was fastest and much more its standard deviation was small. Various factors which may affect safe speed were also described.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.171-178
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• 2014

After the expiration of operating license of a plant, all infrastructures within the plant must be safely dismantled to the point that it no longer requires measures for radiation protection. Despite the fact that Kori 1 and Wolsong 1 are close to the expiration of their operating license, sufficient technologies for radiological characterization, decontamination and dismantling is still under development. The purpose of this study is to develop one of methods for radiological inventory assessment on measuring object by using direct measure of large component with In-Situ measurement technique. Radiological inventory was assessed by analyzing nuclide using portable gamma spectroscopy without dismantling reactor head, and the result of direct measurement was supplemented by performing indirect measurement. Radiochemical analysis were performed on surface contamination samples as well. During the study, radiological inventory of reactor vessel calculated expanding the result. Based on the result and the radioactivity variation of each radionuclides time frame for decommissioning can be decided. Thus, it is expected that during the decommissioning of plants, the result of this study will contribute to the reduction of radiation exposure to workers.