• Proceedings of the Korea Information Processing Society Conference
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• 2013.05a
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• pp.277-280
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• 2013

원자로 건물 내부에는 원자로의 안전운전을 위한 다수의 계통, 구조물 및 기기들이 위치하고 있다. 원자로의 안전과 직결되는 안전계통 변수 (수위, 압력) 들은 경우에 따라서는 현장확인이 요구된다. 원자로 가동 중에 운전원에 의한 안전계통변수의 현장확인이 용이하도록 원자로 건물 내부의 출입구 (2 중문) 부근에 수위계, 압력계 등의 현장계기가 위치하고 있다. 본 논문에서는 일본의 (주) 동경전력이 공개한 후쿠시마 제 1 원자력발전소 1 호기 원자로건물의 IC (비상용 복수기, isolation condenser) 조사영상에 나타나는 현장계기의 영상인식에 대해 기술한다. 조사 영상의 분석에 의하면 현장계기들은 기계식의 아날로그 타입이다. 아날로그 타입의 기하구조를 이용하여 계기판 눈금을 인식하고, 계기판 바늘의 기울기 계산을 통해 계기판을 영상 판독할 수 있었다. 이러한 계기판의 영상판독은 후쿠시마 원전 사고와 같이, 고방사선 피폭 우려로 인해 사람대신에 로봇이 원자로 건물내부에 진입하여 주요 계통, 구조물 및 기기의 현장계기를 판독한다고 가정하면, 유용한 기능이다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.437-445
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• 2019

After the permanent shutdown of K1 in 2017, decommissioning processes have attracted great attention. According to the current decommissioning roadmap, the dismantling of the activated components of K1 may start in 2026, following the removal of its spent fuel. Since the reactor vessel (RV) and reactor vessel internal (RVI) of K1 contain massive components and are relatively highly activated, their decommissioning process should be conducted carefully in terms of radiological and industrial safety. For achieving maximum efficiency of nuclear waste management processes for K1, we present activation analysis of the segmentation process and waste classification of the RV and RVI components of K1. For RVI, the active fuel regions and some parts of the upper and lower active regions are classified as intermediate-level waste (ILW), while other components are classified as low-level waste (LLW). Due to the RVI's complex structure and high activation, we suggest various underwater segmentation techniques which are expected to reduce radiation exposure and generate approximately nine ILW and nineteen very low level waste (VLLW)/LLW packages. For RV, the active fuel region and other components are classified as LLW, VLLW, and clearance waste (CW). In this case, we suggest in-situ remote segmentation in air, which is expected to generate approximately forty-two VLLW/LLW packages.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.291-303
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• 2020

The Kori-Unit 1 nuclear power plant, which is scheduled to be decommissioned after permanent shutdown, is expected to generate large amounts of various types of radioactive waste during the decommissioning process. Among these, nuclear reactors and internal structures have high levels of radioactivity and the dismantled structure must have the proper size and weight on the primary side. During decommissioning, it is important to prepare an appropriate and efficient disposal method through analysis of the disposal status and the legal restrictions on wastes generated from the reactors and internal structures. Nuclear reactors and internal structures generate radioactive wastes of various levels, such as medium, very low, and clearance. A radiation evaluation indicates that wastes in the clearance level are generated in the reactor head and upper head insulation. In this study, a clearance waste safety evaluation was conducted using the RESRAD-RECYCLE code, which is a safety evaluation code, based on the activation evaluation results for the clearance level wastes. The clearance scenario for the target radioactive waste was selected and the maximum individual and collective exposure doses at the time of clearance were calculated to determine whether the clearance criteria limit prescribed by the Nuclear Safety Act was satisfied. The evaluation results indicated that the doses were significantly low, and the clearance criteria were satisfied. Based on the safety assessment results, an appropriate metal recycle and disposal method were suggested for clearance, which are the subject of the deregulation of internal structures of nuclear power plant.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.9
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• pp.1127-1132
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• 2013

The failure of reactor internals may have a significant effect on the safe operation and shutdown of a reactor. Various agings related to neutron irradiation occur or can potentially occur in the reactor internals owing to high neutron irradiation levels. Austenitic stainless steel, one of the principal materials constituting the reactor internals, shows different mechanical material behaviors such as tensile/creep properties and fracture toughness with neutron irradiation levels. This variation should be considered when the structural integrity of the reactor internals against agings during the design lifetime or continued operation period is evaluated. In this study, user subroutine programs considering the variation of mechanical material behaviors with neutron irradiation levels were developed. The programs were validated by testing them for various conditions.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.04a
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• pp.69-70
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• 2014

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.21 no.1
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• pp.47-55
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• 2011

Reactor vessel internals comprehensive vibration assessment program(RVI CVAP) is one of the necessary tests to ensure the safety of nuclear power plants. RVI CVAP of U.S. nuclear regulatory commission regulatory guide 1.20(U.S. NRC R.G. 1.20) consists of the analysis, measurement and inspection. One of the core technologies of the measurement program for RVI CVAP is to select suitable sensors because the measurement is conducted during the critical path of the construction period of nuclear power plants. Therefore, we analyzed RVI thermal-hydraulic and structure design data of Palo Verde nuclear power plant(U.S.), Yonggwang nuclear power plant(Korea) and APR1400 and researched measuring sensors used in them; moreover, we investigated sensors used for measurement of RVI CVAP for the last 20 years throughout the world. Based on these results, we selected suitable measuring sensors for RVI CVAP in advanced power reactor 1400(APR1400).

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.2
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• pp.50-57
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• 2016

The study presents structural integrity assessment of ductility degradation of a baffle former assembly by performing finite element analysis considering real loading conditions and stress triaxiality. Variations of fracture strain curves of type 304 austenitic stainless steel with stress triaxiality are derived based on the previous study results. Temperature distributions during normal operation such as heat-up, steady state, and cool-down are calculated via finite element temperature analysis considering gamma heating and heat convection with reactor coolant. Variations of stress and strain state during long operation period are also calculated by performing sequentially coupled temperature-stress analysis. Fracture strain is derived by using the fracture curve and the stress triaxility. Finally, variations of ductility degradation damage indicator with the fracture strain and the equivalent inelastic strain are investigated. It is found that maximum value of the ductility degradation damage index continuously increases and becomes 0.4877 at 40 EFPYs. Also, the maximum value occurs at top and middle inner parts of the baffle former assembly before and after 20 EFPYs, respectively.

• Computational Structural Engineering
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• v.9 no.1
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• pp.47-54
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• 1996

The Control Element Assembly(CEA) shroud is one of the most important components in the reactor vessel internals for the nuclear power plant. Because of the severe modification from its original design the structural integrity of this component has been questioned. In an attempt to resolve this question, the response of the CEA shroud to a random loading in the actual operating condition is calculated analytically and experimentally and compared to the code allowables to show that it is structurally adequate and acceptable for the long term operation.

• Journal of KSNVE
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• v.21 no.4
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• pp.4-12
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• 2011

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.306-311
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• 1995

원자로 내부구조물의 진동감시를 위한 신호로 노외 중성자 잡음을 선정하고 중성자 잡음에 실려 있는 진동모우드 성분을 신호간의 위상차에 근거하여 분리하는 새로운 방법을 제안하였다. 이 방법은 기존의 방법에 비해 계산속도가 빠르고 3개 이상의 신호에도 적용할 수 있다. 제안한 방법을 토대로 울진 1호기의 노외 중성자 잡음 신호를 채취, 분석하여 내부구조물의 진동특성을 밝히고 진동 감시의 대상이 되는 진동모우드를 분리하므로써 방법의 효율성을 검증하였다. 또 한 핵연료주기동안 주기적으로 신호를 채취, 분석하여 중성자 잡음신호의 특성이 한 핵주기동안 점차 증가하는 경향을 갖는다는 것을 밝혔다.