• 한국경영과학회:학술대회논문집
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• 한국경영과학회 2003년도 추계학술대회 및 정기총회
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• pp.354-357
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• 2003

In decommissioning activities of nuclear facilities, the analyses and evaluations of data(man power, radiation exposure, waste outputs, equipments, etc ) is very important to accomplish effective decommissioning and dismantling. To handle the data from decommissioning activities of research reactor, we developed its database system. By using database system, we come to filter and analyze the date of decommissioning activities and we can evaluate the extraordinary decommissioning features of research reactor. It is likely to lay out and estimate the optimal plan of decommissioning. We expect that the database system of effective decommissioning activities build up the foundation of commercial reactor decommissioning.

• 한국기계가공학회지
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• 제21권1호
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• pp.22-27
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• 2022

In accordance with the growing trend of decommissioning nuclear facilities, research on the cutting process is actively proceeding worldwide. In general, a thermal cutting process, such as plasma cutting is applied to decommissioning a nuclear reactor pressure vessel (RPV). Plasma cutting has the advantage of removing the radioactive materials and being able to cut thick materials. However, when operating under water, the molten metal remains in the cut plane and re-solidifies. Hence, cutting is not entirely accomplished. For these environmental reasons, it is difficult to cut thick metal. The contact arc metal cutting (CAMC) process can be used to cut thick metal under water. CAMC is a process that cuts metal using a plate-shaped electrode based on a high-current arc plasma heat source. During the cutting process, high-pressure water is sprayed from the electrode to remove the molten metal, known as rinsing. As the CAMC is conducted without using a shielding gas, such as Argon, the electrode is consumed during the process. In this study, CAMC is introduced as a method for dismantling nuclear vessels and the relationship between the metal removal and electrode consumption is investigated according to the cutting conditions.

• 시스템엔지니어링학술지
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• 제14권1호
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• pp.63-71
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• 2018

This paper discusses decommissioning procedure requirements management using requirement engineering to systematically manage the technical requirements and criteria that are required in decontamination and decommissioning activities, and the regulatory requirements that should be complied with in a decommissioning strategy for research reactors and nuclear power plants. A schema was designed to establish the traceability and change management related to the linkage between the regulatory requirements and technical criteria after classifying the procedures into four groups during the full life-cycle of the decommissioning. The results confirmed that the designed schema was successfully traced in accordance with the regulatory requirements and technical criteria required by various fields in terms of decontamination and decommissioning activities. In addition, the changes before and after the revision of the Nuclear Safety Act were also determined. The dismantling procedure requirement management system secured through this study is expected to be a useful tool in the integrated management of radioactive waste, as well as in the dismantling of research reactor and nuclear facilities.

• 디지털융복합연구
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• 제13권4호
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• pp.303-310
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• 2015

최근 게임 기술은 모의 수술 시뮬레이션이나 사이버 모델하우스 구축 등의 여러 응용 분야에까지 적용되어지고 있다. 이러한 응용 분야에서 꼭 필요하고 중요한 기술 중 하나는 3D 모델을 실시간으로 절단하는 것이다. 실시간 모의 절단 기술은 자동차나 건축물의 실시간 해체 시뮬레이션 구축에 꼭 필요하며, 다양한 융복합 산업 분야에서 응용이 가능하다. 기존의 BSP-Tree를 이용한 절단 기법에서는 무한 평면을 기준으로 3D 모델을 두 부분으로 절단시키기 때문에 일반적인 절단 작업에 유용하게 사용하기 어렵다. 본 논문에서는 이러한 문제점을 해결하기 위해 유한한 영역 내에서 3D 모델을 절단하는 기법을 제안하였다. 구체적으로 절단 경로면을 유한하게 정의할 수 있도록 하였으며, 절단 범위 이내에서만 3D 모델을 분열시키도록 하여, 다양한 산업 분야에서 유용하게 사용할 수 있도록 하였다. 본 연구에서 제안한 부분 절단 기법의 유용성을 보이기 위해 원자로 3D 모델의 해체 작업 과정을 모의 시뮬레이션 하는 과정에 적용해 보았다.

• Journal of Radiation Protection and Research
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• 제24권2호
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• pp.109-120
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• 1999

TRIGA Mark-II, III 연구로를 폐로함에 따라, 원자로를 비롯한 관련 시설로부터 작업자의 방사선피폭을 최소화하고 환경으로의 방사성물질 누출을 사전에 방지하며, 안전한 해체방법을 모색하여 해체공사시의 기술기준을 수립하기 위하여 해체대상 시설내에 잔존하고 있는 방사성물질의 방사선/능 준위를 측정 및 분석 평가하였다. 측정대상은 시설내의 바닥 및 벽면과 잔존 실험시설물의 표면, 수조내 방사화 물질의 표면, 시설내 잔존하고 있는 냉각수 등이다. 이들 측정대상에 대한 방사선량율과 표면오염도, 핵종 등을 측정, 분석 및 평가하였다. 또한 전산코드 Fispin 을 사용하여 원자로 수조내의 주요 방사화 물질에 대한 방사능양과 핵종을 평가하였다. 본 평가 결과는 해체계획서 작성시 기본자료로서 유용하게 사용되었다.

• 한국환경과학회지
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• 제15권3호
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• pp.211-219
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• 2006

An experimental study on the removal of VOCs gas using a biomembrane reactor were carried out at various inlet gas concentration, specific loading rate, retention time and gas flow rate of volume. The variations of efficiency and various parameters, which are relevant to gas removal, with mixing of soluble gas and without have been discussed. More than 95% of the toluene and methanol present in the feed was successfully removed in each study. The elimination of methanol with mixture of soluble compound of about 300 mg/h corresponds to a portion of 21% if there is a feed stream of 1400 mg/h. On the contrary the maximum efficiency of about 72% of toluene was reached. This is to be rated as a treatment of sorption that the limiting factor of the dismantling speed could be represented by this difficult degradable component. Nevertheless the elimination capacities for this reactor for toluene were on a very high level. For substances which show a very high solubility in silicon rubber an advantage of a bio membrane is clearly shown. Therefore a similarly good result is expected for n-hexane, because of its relatively good permeability which was distinguished during permeation experiments.

• 한국압력기기공학회 논문집
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• 제16권2호
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• pp.18-24
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• 2020

When a damage occurs in the nuclear fuel burning in the reactor, fission products that should be in the nuclear fuel rod are released into the reactor coolant. In this case, sipping test, a series of non-destructive inspection methods, are used to find leakage in nuclear fuel assemblies during the power plant overhaul period. In addition, the sipping test is also used to check the integrity of the spent fuel for moving to an intermediate dry storage, which is carried out as the first step of nuclear decommissioning, . In this paper, the principle and characteristics of the sipping test are described. The structure of the sipping inspection equipment is largely divided into a suction device that collects fissile material emitted from a damaged assembly and an analysis device that analyzes their nuclides. In order to make good use of the sipping technology, the radioactive level behavior of the primary system coolant and major damage mechanisms in the event of nuclear fuel damage are also introduced. This will be a reference for selecting an appropriate sipping method when dismantling a nuclear power plant in the future.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2004년도 Proceedings of the 4th Korea-China Joint Workshop on Nuclear Waste Management
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• pp.96-105
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• 2004

The decommissioning project of the KRR 1 & 2 was started in January 1997. The actual decommissioning activity was started at the RI production facility and was finished at the end of 2002. The dismantling works of all components including the reactor structure of the KRR-2 was started in January, 2003 and will be carried out for 2 years till the end of 2004. The project schedule is estimated to delay for 4∼5 months beyond the original plan because of delaying on the cutting of thermal column nose and removal of the graphite bricks, but it may be caught up during the removal working of concrete from biological shielding structure. This paper summarizes the general status of the KRR 1 & 2 and decommissioning activities.

• Nuclear Engineering and Technology
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• 제55권8호
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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.

• 방사선산업학회지
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• 제18권2호
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• pp.141-145
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• 2024

In Publication No. 66 of the International Commission on Radiological Protection, an activity median aerodynamic diameter (AMAD) of 5 ㎛ is considered in internal exposure dose assessment owing to inhalation of radionuclides in a workplace. However, analysis of aerosols generated during dismantling experiments, such as in the oxy-cutting of a reactor vessel conducted in Korea, revealed that the radioactive aerosols have AMAD ranging from 0.024 to 0.064 ㎛. Such extremely fine aerosols can induce internal exposure if inhaled. In particular, alpha radionuclides in aerosols can lead to significantly higher levels of radiation exposure than beta and gamma radionuclides, thus highlighting the need to establish appropriate internal exposure radiation protection programs and monitoring systems that specifically address alpha radionuclides when decommissioning nuclear power plants in Korea.