• Journal of the Korean Society of Safety
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• v.36 no.5
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• pp.86-91
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• 2021

Calculating the scrutable core damage frequency (CDF) of nuclear power plants is an important component of the seismic probabilistic safety assessment (SPSA). In this work, a simple approach is developed to calculate CDF from minimal cut sets (MCSs) with non-rare events. When conventional calculation methods based on rare event approximations are employed, the CDF of industry SPSA models is significantly overestimated by non-rare events in the MCSs. Recently, quantification algorithms using binary decision diagrams (BDDs) have been introduced to prevent CDF overestimation in the SPSA. However, BDD structures are generated from a small part of whole MCSs due to limited computational memory, and they cannot be reviewed due to their complicated logic structure. This study suggests a simple approach for scrutinizing the CDF calculation based on whole MCSs in the SPSA system analysis model. The proposed approach compares the new results to outputs from existing algorithms, which helps in avoiding CDF overestimation.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2022.05a
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• pp.566-568
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• 2022

Metaverse is the most emerging technology due to the recent 4th industry and the non-face-to-face society of Corona 19. As one of the core technologies of Metaverse, VR·AR technology is being industrialized in various fields such as medical care, education, and service. Among them, education and training are the most important fields of application, and nuclear power plant operation also requires this technology. In this paper, we will look at the fields of application of VR·AR technology in existing industries and suggest a plan for use in nuclear power plant work.

• Nuclear Engineering and Technology
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• v.42 no.5
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• pp.590-599
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• 2010

Recently, efficient operation and practical management of power plants have become important issues in the nuclear industry. In particular, typical aging parameters such as stress and cumulative usage factor should be determined accurately for continued operation of a nuclear power plant beyond design life. However, most of the major components have been designed via conservative codes based on a 2-D concept, which do not take into account exact boundary conditions and asymmetric geometries. The present paper aims to suggest an effective fatigue evaluation methodology that uses a prototype of the integrated model and its transfer functions. The validity of the integrated 3-D Finite Element (FE) model was proven by comparing the analysis results of individual FE models. Also, mechanical and thermal transfer functions, known as Green's functions, were developed for the integrated model with the standard step input. Finally, the stresses estimated from the transfer functions were compared with those obtained from detailed 3-D FE analyses results at critical locations of the major components. The usefulness of the proposed fatigue evaluation methodology can be maximized by combining it with an on-line monitoring system, and this combination, will enhance the continued operations of old nuclear power plants.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.532-548
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• 2021

Nuclear power plants contain several monitoring systems that can identify the in-vessel phenomena of a severe accident (SA). Though a lot of analysis and research is carried out on SA, right from the development of the nuclear industry, not all the possible circumstances are taken into consideration. Therefore, to improve the efficacy of the safety of nuclear power plants, additional analytical studies are needed that can directly monitor severe accident phenomena. This paper presents an interacting multiple model (IMM) based fault detection and diagnosis (FDD) approach for the identification of in-vessel phenomena to provide the accident propagation information using reactor vessel (RV) out-wall temperature distribution during severe accidents in a nuclear power plant. The estimation of wall temperature is treated as a state estimation problem where the time-varying wall temperature is estimated using IMM employing three multiple models for temperature evolution. From the estimated RV out-wall temperature and rate of temperature, the in-vessel phenomena are identified such as core meltdown, corium relocation, reactor vessel damage, reflooding, etc. We tested the proposed method with five different types of SA scenarios and the results show that the proposed method has estimated the outer wall temperature with good accuracy.

• Corrosion Science and Technology
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• v.10 no.2
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• pp.60-64
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• 2011

Lean duplex stainless steels have been developed in Korea for the purpose of being used in the seawater systems of industry. There are also many important seawater systems in nuclear power plants. These systems supply seawater to cooling water condenser tubes, heat exchanger tubes, related pipes and chlorine injection systems. The flow velocity of some part of seawater systems in nuclear power plants is high and damages of components from corrosion are severe. The considered lean duplex stainless steels are STS329LD (20.3Cr-2.2Ni-1.4Mo) and STS329J3L (22.4Cr-5.7Ni-3Mo) and PRENs of them are 29.4 and 37.3 respectively. Physical, mechanical and micro-structural properties of them are evaluated, and electrochemical corrosion resistance is measured quantitatively in NaCl solution. Critical Pitting Temperatures (CPT)s are measured on these alloys and pit depths are evaluated using laser microscope. Long period field tests on these alloys are now being performed, and some results are going to be presented in the following study.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2812-2820
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• 2020

House load operation (HLO) occurs when the generator supplies power to the house load without triggering reactor trips during grid disturbances. In Korea, the HLO capability of optimized power reactor 1000 (OPR1000) plants has prevented several reactor trips. Operational experiences demonstrate the difference in the reactor trip incidence due to grid disturbances between OPR1000 plants and Westinghouse plants in Korea, attributable to the availability of the HLO capability. However, probabilistic safety assessments (PSAs) for OPR1000 plants have not considered their specific design features in the initiating event analyses. In an at-power PSA, the HLO capability can affect the initiating event frequencies of general transients (GTRN) and loss of offsite power (LOOP), resulting from transients within the grid system. The initiating event frequencies of GTRN and LOOP for an OPR1000 plant are reduced by 17.7% and 78.7%, respectively, compared to the Korean industry-average initiating event frequencies, and its core damage frequency from internal events is reduced by 15.2%. The explicit consideration of the HLO capability in initiating event analyses makes significant changes in the risk contributions of the initiating events. Consequently, for more realistic at-power PSAs in Korea, we recommend incorporating plant-specific HLO-related design features when estimating initiating event frequencies.

• Journal of the Korean Professional Engineers Association
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• v.32 no.3
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• pp.62-71
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• 1999

• Journal of the Korean Professional Engineers Association
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• v.32 no.5
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• pp.77-84
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• 1999

• Journal of the Korean Professional Engineers Association
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• v.32 no.4
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• pp.57-63
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• 1999

• JOURNAL OF ELECTRICAL WORLD
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• no.2 s.158
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• pp.5-9
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• 1990