• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1234-1245
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• 2018

The risk of multi-unit nuclear power plants (NPPs) at a site has received considerable critical attention recently. However, current probabilistic safety assessment (PSA) procedures and computer code do not support multi-unit PSA because the traditional PSA structure is mostly used for the quantification of single-unit NPP risk. In this study, the main purpose is to develop a multi-unit Level 2 PSA method and apply it to full-power operating six-unit OPR1000. Multi-unit Level 2 PSA method consists of three steps: (1) development of single-unit Level 2 PSA; (2) extracting the mapping data from plant damage state to source term category; and (3) combining multi-unit Level 1 PSA results and mapping fractions. By applying developed multi-unit Level 2 PSA method into six-unit OPR1000, site containment failure probabilities in case of loss of ultimate heat sink, loss of off-site power, tsunami, and seismic event were quantified.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.2606-2611
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• 2008

The high energy fluid leakage from the high temperature and high differential pressure drop system of NPPs (Nuclear Power Plants) decreases efficiency and consequently leads to considerable economic loss due to less power production. Also, the leakage possibly damages critical parts of components such as valve and trim with the effect of cavitation, flashing, and erosion, etc. and deteriorates its performance. Thus, in this study, we diagnosed the as-is leakage for four (4) main steam drain valves and two (2) steam traps of Yonggwang 1,2 units during normal operation by using multi-measuring technique and observed the occurrence of fine leakage. In the course of measuring fluid leakage, the sign of fine leakage is estimated to be the leakage from orifice. By converting the leakage to energy loss, it is equivalent to the amount of several hundred thousand won per each unit, which supports the basis for the justification of fine leakage.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.1-7
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• 2013

In around 2000, the U.S. NPPs have developed the various advanced engineering processes based on the INPO AP-913(Equipment Reliability Process Description) and showed the high performance in availability. With these benchmarking cases, the Korean NPPs have introduced the advanced engineering technology since 2005. The first step of the advanced engineering is to analyze and determine component importance for all components of a plant. This process is called Functional Importance Determination(FID). These results are basically utilized to determine the priority with limited resources in various areas. However, because the consistency of FID results is insufficient despite applying the same criteria in the existing operating NPPs, the degree of application is low. Therefore, this paper presents the improved methodology for FID interfacing system functions of Maintenance Rule Program and results of Single Point Vulnerability(SPV). This improved methodology is expected to contribute to enhance the reliability of FID data.

• Journal of Energy Engineering
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• v.23 no.2
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• pp.183-190
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• 2014

A Commercial Grade Item Dedication(CGID) for Korean operating nuclear power plants has been implemented since 2012. The CGID implementation and strategies for Korea are established as follows: CGID policy establishment, R&D of a specific methodologies of CGID, enrollment of third party organizations for CGID work, CGID program establishment for enrolled suppliers, establishment of training courses for certification, and CGID process development for quality class Q and A. Consequently, it is expected that these activities are enable to enhance the reliability and the safety of components and/or parts in nuclear power plants.

• Nuclear Engineering and Technology
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• v.36 no.6
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• pp.485-496
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• 2004

Typical probabilistic safety assessment (PSA) results were used to estimate the performance indicator (PI) thresholds of unplanned reactor scram (URS) and safety system unavailability (SSU) for Korean nuclear power plants (NPPs). The changes in core damage frequency (${\Delta}$CDFs) of $10^{-6}/yr$, $10^{-5}/yr$, and $10^{-4}/yr$ were adopted as the risk criteria in setting up the PI thresholds. The PI thresholds for the URS were estimated using information pertaining to the initiating event frequencies, the CDF, and the CDF contribution of each initiating event. The PI thresholds of the SSU were estimated using information on the unavailability, the Fussell-Vesely importance, and the CDF.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.27 no.5
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• pp.385-392
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• 2014

The isolators have been generally used to reduce a seismic force. If the isolators apply to the nuclear power plants(NPPs), the durability and capacity for the structures and equipments should be ensured during the life time. In this study, the long-term behavior of isolated NPPs was analyzed for ensuring the seismic safety. The properties of isolator due to the age-related degradation were analyzed. And the seismic behavior of isolated buildings was analyzed by considering the aging of rubber bearings in different temperature environments. According to the analysis results, the natural frequency of structures was increased with time. But the maximum acceleration and maximum displacement of isolated structures have not changed significantly. Although the damaged of structure did not occurred by aging of isolators, it was presented that the spectral acceleration at the target frequency of isolated structure increased with the temperature. Therefore the isolators in the isolated buildings should be carefully designed and manufactured considering the temperature-dependancy of rubber material.

• Proceedings of the Korean Reliability Society Conference
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• 2000.11a
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• pp.257-267
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• 2000

Mostly, the economic analyses for replacement of major components of nuclear power Plants(NPPs) have been performed in deterministic ways. However, the analysis results are more or less affected by the uncertainties associated with input variables. Therefore, it is desirable to use a probabilistic economic analysis method to properly consider uncertainty of real problem. In this paper, the probabilistic economic analysis method and decision analysis technique are briefly described. The probabilistic economy analysis method using decision analysis will provide efficient and accurate way of economic analysis for the repair and/or replace mai or components of NPPs.

• Journal of the Korean Society of Safety
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• v.38 no.3
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• pp.102-109
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• 2023

Following the accident at Fukushima, the true impact of multi-unit accidents came to light. Accordingly, research related to multi-unit accident effect analysis, risk evaluation, and accident prevention/prevention technology has been conducted. Specific examples are mobile/fixed equipment such as multi-barrier accident coping strategy (MACST) and diverse and flexible coping strategies (FLEX), which have been introduced and installed in multi-units for preventing and mitigating multi-unit accidents. These strategies are useful for enhancing the safety of nuclear power plants (NPPs); however, a more efficient strategy is required in terms of the costs of physical and human resources. To effectively and efficiently mitigate an increase in multi-unit accidents, it is necessary to not only to utilize mobile/fixed equipment but to also use crosstie options with resources that already exist at NPPs. Therefore, we analyzed the current international and domestic status of crosstie systems technology and propose a method to evaluate feasibility alongside risk based on a multi-unit probabilistic safety assessment (PSA). To analyze the international and domestic status of crosstie systems technology, actual cases and related research were studied, and a list of potential crosstie safety resources was derived. Additionally, a case study was performed on crosstie cases of two systems within the assumed six units on-site under a multi-unit accident, and a multi-unit PSA-based risk evaluation method is proposed.

• Proceedings of the KSME Conference
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• 2007.05a
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• pp.1702-1707
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• 2007

The reduction of pipe-thickness induced by flow accelerated corrosion (FAC) is one of the most serious problems on the maintenance of piping system in nuclear power plants (NNP). If the thickness of a pipe component is reduced below the critical level, it cannot sustain pressure and consequently results in leakage or rupture. For this reason, wall thinning by FAC has been inspected in secondary side piping systems in NPPs. In this research Round Robin Test (RRT) was conducted to verify confidence of wall thinning measurement system in NPP. 12 inspectors from 3 companies participated and 23 specimens were used according to standard practice in RRT. The gage R&R analysis was introduced in regard to repeatability and reproducibility that are affected to measurement system errors. Confidence intervals of thickness measurement system were obtained.

• Journal of Magnetics
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• v.16 no.2
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• pp.88-91
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• 2011

Steam generator tubes (SGTs) in nuclear power plants (NPPs) are a boundary between the primary side generating heat by nuclear fission and the secondary side generating electric power by a turbine. The water inside the SGT is high temperature and high pressure. Therefore, defects and magnetic phases (MPs) are partly produced in non-magnetic SGT by high stresses and temperatures. This causes trouble regarding the safety of SGTs but it is difficult to detect the MP using the conventional eddy current technique (ECT). In particular, a circumferential defect (CD) and circumferential magnetic phase (CMP) cannot detected by ECT. Consequently, a new method is needed to detect CDs and CMPs in SGT. A new U-type yoke with two types of coils was designed and the reactance signal by the CMPs and CDs in the SGT material was simulated.