• Nuclear Engineering and Technology
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• 제52권1호
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• pp.59-68
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• 2020

ECCMIX component was introduced in RELAP5/MOD3 for calculating the interfacial condensation. Compared to other existing components in RELAP5, user experience of ECCMIX component is restricted to developmental assessment applications. To evaluate the capability of the ECCMIX component, ECCS experiment was conducted which included single-phase and two-phase thermal mixing. The experiment was carried out with test sections containing a main pipe (70 mm inner diameter) and a branch pipe (21 mm inner diameter) under the atmospheric pressure. The steam mass flow in the main pipe ranged from 0 to 0.0347 kg/s, and the subcooled water mass flow in the branch pipe ranged from 0.0278 to 0.1389 kg/s. The comparison of the experimental data with the calculation results illuminated that although the ECCMIX component was more difficult to converge than Branch component, it was a more appropriate manner to simulate interfacial condensation under two-phase thermal mixing circumstance, while the two components had no differences under single-phase circumstance.

• 시스템엔지니어링학술지
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• 제12권2호
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• pp.69-80
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• 2016

The purpose of this study is to establish a component reutilization system in Nuclear Power Plant (NPP) by Object-Oriented Systems Engineering Method (OOSEM). Unified Modeling Language (UML) is mainly used for OOSEM. Operational Concept (OpsCon), Use cases, Structure Diagrams, and Behavior Diagrams are developed to analyze stakeholders needs, system requirements, logical architecture, and physical architecture. Based on the current decommissioning and purchasing system of the component, some activities from their systems were excepted and additional new activities were developed for a component reutilization system.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2974-2988
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• 2022

Implementation of component isolation in nuclear industry is challenging due to gaps in research and the lack of specific guidelines. In this study, parameters affecting component-level isolation of advanced reactor vessels are identified based on a representative numerical model with explicit consideration of nonlinear soil-structure interaction (SSI). The objective of this study is to evaluate the effectiveness of, and to identify potential limitations of using conventional friction pendulum bearings to seismically isolate vessels. It is found that slender vessels or components are particularly vulnerable to rotational accelerations at the isolation interface, which are caused by rotation of the sub-structure and by excitation of higher modes in the horizontal direction of the seismically isolated system. Component isolation is found to be more effective for relatively stiffer vessels and at sites with stiff soil. Considering that conventional isolators are deficient in resisting axial tension, it is observed that the optimum location for supporting a component to achieve seismic isolation, is at a cross-sectional plane passing through the center of mass of the vessel. These findings are corroborated by numerous simulations of the response of seismically isolated reactor vessels at different nuclear power plant sites subject to a variety of ground motions.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.175.1-175
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• 2002

• 에너지공학
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• 제20권2호
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• pp.154-162
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• 2011

예방정비기준은 기기의 높은 신뢰성을 유지하기 위해 정비방법을 문서화한 것으로 기기별 최적화된 예방정비 항목과 주기를 명시하고 기기의 성능저하 메커니즘을 설명하는 지침서이다. 기존의 기기 또는 설비에 대한 정비전략은 상태감시 정비보다는 일정한 주기가 되면 정비를 수행하는 주기정비가 주된 전략이었다. 본 논문은 원전 정지형 여자기에 대하여 기존의 주기정비 형태가 아닌 상태기반 정비 기법을 강화하고 정비 프로그램을 표준화하여 최적화된 정비전략을 구축하기 위한 예방정비 기준을 개발한 내용으로 기기 신뢰성, 발전소 이용률을 향상시킬 수 있는 기반이 될 것으로 판단된다.

• 시스템엔지니어링학술지
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• 제16권1호
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• pp.84-97
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• 2020

This paper aims to develop an optimized inventory management method for safety critical Instrument and Control (I&C) components. In this regard, the paper focuses on estimating the consumption rate of I&C components using demand forecasting methods. The target component for this paper is the Foxboro SPEC-200 controller. This component was chosen because it has highest consumption rate among the safety critical I&C components in Korean OPR-1000 NPPs. Three analytical methods were chosen in order to develop the demand forecasting methods; Poisson, Generalized Linear Model (GLM) and Bootstrapping. The results show that the GLM gives better accuracy than the other analytical methods. This is because the GLM considers the maintenance level of the component by discriminating between corrective and preventive.

• Nuclear Engineering and Technology
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• 제51권3호
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• pp.709-718
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• 2019

The international nuclear industry has undergone a lot of changes since the Fukushima, Chernobyl and TMI nuclear power plant accidents. However, there are still large and small component deficiencies at nuclear power plants in the world. There are many causes of electrical equipment defects. There are also factors that cause component failures due to human errors. This paper analyzed the root causes of failure and types of human error in 300 cases of electrical component failures. We analyzed the operating experience of electrical components by methods of root causes in K-HPES (Korean-version of Human Performance Enhancement System) and by methods of human error types in HuRAM+ (Human error-Related event root cause Analysis Method Plus). As a result of analysis, the most electrical component failures appeared as circuit breakers and emergency generators. The major causes of failure showed deterioration and contact failure of electrical components by human error of operations management. The causes of direct failure were due to aged components. Types of human error affecting the causes of electrical equipment failure are as follows. The human error type group I showed that errors of commission (EOC) were 97%, the human error type group II showed that slip/lapse errors were 74%, and the human error type group III showed that latent errors were 95%. This paper is meaningful in that we have approached the causes of electrical equipment failures from a comprehensive human error perspective and found a countermeasure against the root cause. This study will help human performance enhancement in nuclear power plants. However, this paper has done a lot of research on improving human performance in the maintenance field rather than in the design and construction stages. In the future, continuous research on types of human error and prevention measures in the design and construction sector will be required.

• 한국안전학회지
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• 제26권6호
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• pp.85-89
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• 2011

Component reliability data were analyzed by using PRinS(Plant Reliability data information System) based on the latest operating experiences of eight KSNPs(Korea Standard Nuclear Power plants), and these new data were applied to the KSNP PSA models. In addition, the existing PSA models were revised for reflecting as-built and as-operated plant conditions. As a result of newly performing PSA in this paper, CDF and LERF were estimated 26.1% and 18.2% lower than the existing values, respectively. It was identified that the risk measures decreased not because of revising the models but because of applying the new component reliability data. The result and the method of this paper could be used when generating plant specific data and performing the living PSA in the future.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.526-535
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• 2024

This paper proposes a safety-critical standby component unavailability model that contains aging effects caused by the elapsed time from installation, component degradation due to surveillance tests, and imperfect maintenance actions. An application of the model to a Motor-Operated Valve and a Motor-Driven Pump involved in the HPIS of a VVER/1000-V446 nuclear power plant is demonstrated and compared with other existing models at component and system levels. In addition, the effects of different unavailability models are reflected in the NPP's risk criterion, i.e., core damage frequency, over five maintenance periods. The results show that, compared with other models that do not simultaneously consider the full effects of degradation and maintenance impacts, the proposed model realistically evaluates the unavailabilities of the safety-related components and the involved systems as a plant age function. Therefore, it can effectively reflect the age-dependent CDF impact of a given testing and maintenance policy in a specified time horizon.

• 한국신뢰성학회:학술대회논문집
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• 한국신뢰성학회 2000년도 추계학술대회
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• pp.269-277
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• 2000

The component reliability database is required in PSA (Probabilistic Safety Analysis) for NPP (Nuclear Power Plant). We have applied a generic database to the PSA for the Korean NPPs, since there is no specific component reliability database. Therefore we are developing the plant-specific component reliability database for domestic NPPs. We also extend the experience and knowledge of PSA and component reliability database for NPP to chemical industry We collect the raw data like component operation history and maintenance history and then input the required data for the component reliability database through failure analysis. With the database, we can not only perform PSA with real data but also perform maintenance optimization.