• Proceedings of the KAIS Fall Conference
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• 2011.12a
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• pp.353-356
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• 2011

자동차 폐차사업자들은 대부분 소규모로 영세하게 운영되고 있으며, 비효율적인 폐차 프로세스로 인해 폐차 소유자는 물론 폐차처리장까지 불이익을 받고 있다. 이를 위한 폐차 시스템에 대한 연구가 국내외에서 활발히 진행되고 있다. 하지만 기존에 개발되어진 폐차 해체시스템의 경우는 국내 폐차업계에 적용하기에는 경제적, 실용적으로 현실성이 부족하다. 따라서 본 연구에서는 폐차에 대한 효율성 극대화와 범용성 확보를 위해 시뮬레이션을 통한 폐자동차 해체시스템을 개발하고자 한다. 시뮬레이션을 통한 폐차 해체시스템의 사용은 복잡한 공정 또는 시스템의 설계 및 운영결과에 대한 효율적 분석이 가능하며, 다양한 위치 및 작업내용을 변경시켜 여러 대안을 마련할 수 있고, 이에 대한 결과를 미리 알아봄으로써 적은 시간과 비용으로 현장상황에 맞는 적절한 시스템을 구축할 수 있게 해주는 장점이 있다. 또한 최적 해체경로 시스템의 적용으로 시스템의 효율을 향상시켰다. 개발되어질 시스템은 높은 범용성 및 유연성 확보로 국내 폐차 업계의 발전에 크게 이바지 할 것이다.

• Journal of Energy Engineering
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• v.16 no.3
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• pp.93-102
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• 2007

This paper presents a scenario evaluation model of the AHP (Analytic Hierarchy Process) to evaluate dismantling scenarios considering quantitative and qualitative considerations. And decommissioning information producing modules which can obtain a dismantling schedule, quantify radioactive waste, visualize a radioactive inventory, estimate a decommissioning cost, and estimate a worker's exposure was developed to assess qualitatively decommissioning information. The digital mock-up (DMU) system was developed to verify dismantling processes and find error of scenarios in virtual space. It combines and manages the decommissioning information producing modules, the decommissioning DB, and the dismantling evaluation module synthetically. By using AHP model and DMU system, the thermal column in KRR-1 was evaluated on plasma arc cutting scenario and nibbler cutting scenario using the developed decommissioning DMU system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.95-106
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• 2019

The decommissioning of nuclear power plants should be prepared by quantitative and qualitative risk assessment. Radiological and non-radiological hazards arising during decommissioning activities must be assessed to ensure the safety of decommissioning workers and the public. Decommissioning experiences by U.S. operators have mainly focused on deterministic risk assessment, which is standardized by the U.S. Nuclear Regulatory commission (NRC) and focuses only on the consequences of risk. However, the International Atomic Energy Agency (IAEA) has suggested an alternative to the deterministic approach, called the risk matrix technique. The risk matrix technique considers both the consequence and likelihood of risk. In this study, decommissioning stages, processes, and activities are organized under a work breakdown structure. Potential accidents in the decommissioning process of NPPs are analyzed using the composite risk matrix to assess both radiological and non-radiological hazards. The levels of risk for all potential accidents considered by U.S. NPP operators who have performed decommissioning were estimated based on their consequences and likelihood of events.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.85-96
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• 2004

This paper is to study on the decommissioning cost estimation for nuclear facilities of advanced nuclear organizations and countries for deriving the cost factors to be taken considerations into accomplishing decommissioning projects. Of cost categories producing the factors of decommissioning costs, dismantling and waste processing & disposals activities are examined to increase the its costs. Of labor, materials and other costs categories, labor costs are summarized to have overall majorities in the decommissioning cost factors. The main parameters of all factors affecting the decommissioning costs are analyzed as work difficulty, regional labor costs, peripheral cost, disposal cost and final burial costs.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.473-478
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• 2018

Radioactive aerosol generated in cutting and melting work during the NPP decommissioning process can cause internal exposure to body through workers' breath. Thus, it is necessary to assess worker internal exposure due to the radioactive aerosol during decommissioning. The actually measured value of the working environment is needed for accurate assessment of internal exposure, but if it is difficult to actually measure that value, the internal exposure dose can be estimated through recommended values such as the fraction of amount of intake and the size of particles suggested by the International Committee on Radiological Protection (ICRP). As for the selection of particle size, this study applied a value of $5{\mu}m$, which is the size of particles considering the worker recommended by the ICRP. As for the amount of generation, the amount of intake was estimated using data on the mass of aerosol generated in a melting facility at a site in Kozloduy, Bulgaria. In addition, using these data, this study calculated the level of radioactivity in the worker's body and stool and conducted an assessment of internal exposure using the BiDAS computer code. The internal exposure dose of Type M was 0.0341 mSv, that of Type S was 0.0909 mSv. The two types of absorption showed levels that were 0.17% and 0.45% of the domestic annual dose limit, respectively.

• 월간산업보건
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• s.229
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• pp.31-33
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• 2007

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2006.06a
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• pp.53-54
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• 2006

• Magazine of RCR
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• v.4 no.1
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• pp.43-48
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• 2009

• 월간산업보건
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• s.230
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• pp.41-44
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• 2007

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2011.10a
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• pp.371-372
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• 2011