• Earthquakes and Structures
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• v.13 no.6
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• pp.561-572
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• 2017

Seismic base isolation has been accepted as one of the most popular design procedures to protect important structures against earthquakes. However, due to lack of information and experimental data the application of base isolation is quite limited to nuclear power plant (NPP) industry. Moreover, the effects of inelastic behavior of soil beneath base-isolated NPP have raised questions to the effectiveness of isolation device. This study applies the wavelet analysis to investigate the effects of soil-structure interaction (SSI) on the seismic response of a base-isolated NPP structure. To evaluate the SSI effects, the NPP structure is modelled as a lumped mass stick model and combined with a soil model using the concept of cone models. The lead rubber bearing (LRB) base isolator is used to adopt the base isolation system. The shear wave velocity of soil is varied to reflect the real rock site conditions of structure. The comparison between seismic performance of isolated structure and non-isolated structure has drawn. The results show that the wavelet analysis proves to be an efficient tool to evaluate the SSI effects on the seismic response of base-isolated structure and the seismic performance of base-isolated NPP is not sensitive to the effects in this case.

• The Transactions of the Korea Information Processing Society
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• v.13 no.5
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• pp.227-235
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• 2024

Since nuclear power plants (NPPs) increasingly employ digital I&C systems, reliability evaluation for NPP software has become crucial for NPP probabilistic risk assessment. Several methods for estimating software reliability have been proposed, but there is no available tool support for those methods. To support NPP software manufacturers, we propose a reliability measurement tool for NPP software. We designed our tool to provide reliability estimation depending on available qualitative and quantitative information that users can offer. We applied the proposed tool to an industrial reactor protection system to evaluate the functionality of this tool. This tool can considerably facilitate the reliability assessment of NPP software.

• Nuclear Engineering and Technology
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• v.36 no.5
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• pp.475-484
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• 2004

Extending the lifetime of a nuclear power plant [(hereafter referred to simply as NPP)] is one of the most important concerns in the global nuclear industry. Cables are one of the long-life items that have not been considered for replacement during the design life of a NPP. To extend the cable life beyond the design life, it is first necessary to prove that the design life is too conservative compared with actual aging. Condition monitoring is useful means of evaluating the aging condition of cable. In order to simulate natural aging in a nuclear power plant. a study on accelerated aging must first be conducted. In this paper, evaluations of mechanical aging degradation for a neoprene cable jacket were performed after accelerated aging under tcontinuous and intermittent heating conditions. Contrary to general expectations, intermittent heating to the neoprene cable jacket showed low aging degradation, 50% break-elongation, and 60% indenter modulus, compared with continuous heating. With a plant maintenance period of 1 month after every 12 or 18 months operation, we can easily deduce that the life time of the cable jacket of neoprene can be extended much longer than extimated through the general EQ test. which adopts continuous accelerated aging for determining cable life. Therefore, a systematic approach that considers the actual environment conditions of the nuclear power plant is required for determining cable life.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.70-74
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• 2014

An Operating Experience Report(OER) has written about events and accidents happened at a Nuclear Power Plant(NPP). The purpose of publishing the OER is to prevent the similar event or accident repeatedly by spreading the experience of a single plant to other plants personnel. In this paper, it is analyses that the foreign NPPs' OERs on JIT published by the International Nuclear Agency(WANO, INPO, COG, BE). The analysis introduced in this paper is performed along with the various factors such as type of work, root-cause, and equipment. The root-cause analysis about the OERs shows that the Human-error is the major factor in foreign NPPs, but on the other hand equipment problem is the main part of the Domestic NPPs. The ratio of the foreign NPP's OERs on JIT according to the type of work was applied to KHNP-JIT developed nowadays for the first time in KOREA.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.176-186
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• 2015

Pattern classifications have become important tools for fault diagnosis in nuclear power plants (NPP). However, it is often difficult to obtain training data under fault conditions to train a supervised classification model. By contrast, normal plant operating data can be easily made available through increased deployment of supervisory, control, and data acquisition systems. Such data can also be used to train classification models to improve the performance of fault diagnosis scheme. In this paper, a fault diagnosis scheme based on semisupervised classification (SSC) scheme is developed. In this scheme, new measurements collected from the plant are integrated with data observed under fault conditions to train the SSC models. The trained models are subsequently applied to new measurements for fault diagnosis. In comparison with supervised classifiers, the proposed scheme requires significantly fewer data collected under fault conditions to train the classifier. The developed scheme has been validated using different fault scenarios on a desktop NPP simulator as well as on a physical NPP simulator using a graph-based SSC algorithm. All the considered faults have been successfully diagnosed. The results have demonstrated that SSC is a promising tool for fault diagnosis in NPPs.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.308-322
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• 2020

This paper presents a conceptual design for a plant-wide autonomous operation system that uses artificial intelligence techniques. The autonomous operation system has the power and ability to perform the control functions needed for the emergency operation of a nuclear power plant (NPP) with reduced operator intervention. This paper discusses the emergency operation and level of automation in an NPP and presents the design requirements for an autonomous emergency operation system (A-EOS). Then, an architecture that consists of several modules is proposed, with descriptions of the functions. Finally, this paper introduces a prototype of the suggested autonomous system that integrates the authors' previous works.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.1115-1125
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• 2022

Nuclear power plant (NPP) is currently considered as one of the most reliable power sources. However, 182 of them are considered decommissioned and inactive including the one in Bataan, Philippines. The aim of this study was to investigate the acceptance of the reopening of Bataan Nuclear Power Plant (BNPP) by integrating the Theory of Planned Behavior and Protection Motivation Theory. A total of 815 Filipinos answered an online questionnaire which consisted of 37 questions. The Structural Equation Modeling (SEM) indicated that knowledge towards nuclear power plants was the key factor in determining people's acceptance towards NPP reopening. In addition, knowing the benefits would lead to positive perceived behavioral control (PBC) and attitude towards intention. Results showed that PBC and attitude are mediators towards the acceptance of people regarding the reopening of BNPP. If an individual's knowledge gravitates towards the perceived risk, then this can lead to the negative acceptance of the NPP reopening. On the other hand, if an individual's knowledge gravitates towards the perceived benefits, then this will lead to positive acceptance. This study is the first study that explored the acceptance of the reopening BNPP. Finally, the study's model construct would also be very beneficial for researchers, government, and even private sectors worldwide.

• Korean System Dynamics Review
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• v.3 no.2
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• pp.49-68
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• 2002

The intent of this study is to develop system dynamics model for assessment of organizational and human factors in nuclear power plant which can contribute to secure the nuclear safety. Previous studies are classified into two major approaches. One is engineering approach such as ergonomics and probability safety assessment(PSA). The other is social science approach such like sociology, organization theory and psychology. Both have contributed to find organization and human factors and to present guideline to lessen human error in NPP. But, since these methodologies assume that relationship among factors is independent they don't explain the interactions among factors or variables in NPP. To overcome these limits, we have developed system dynamics model which can show cause and effect among factors and quantify organizational and human factors. The model we developed is composed of 16 functions of job process in nuclear power, and shows interactions among various factors which affects employees' productivity and job quality. Handling variables such like degree of leadership, adjustment of number of employee, and workload in each department, users can simulate various situations in nuclear power plant in the organization side. Through simulation, user can get insight to improve safety in plants and to find managerial tools in the organization and human side. Analyzing pattern of variables, users can get knowledge of their organization structure, and understand stands of other departments or employees. Ultimately they can build learning organization to secure optimal safety in nuclear power plant.

• Nuclear Engineering and Technology
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• v.34 no.6
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• pp.586-595
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• 2002

The Seismic probabilistic risk assessment (SPRA) or seismic margin assessment (SMA) have been used for the seismic safety evaluation of nuclear power plant structures and equipments. For the SPRA or SMA, the reference response spectrum should be defined. The site-specific median spectrum has been generally used for the seismic fragility analysis of structures and equipments in a Korean nuclear power plant Since the site-specific spectrum has been developed based on the peak ground motion parameter, the site-specific response spectrum does not represent the same probability of exceedance over the entire frequency range of interest. The uniform hazard spectrum is more appropriate to be used in seismic probabilistic risk assessment than the site- specific spectrum. A method for modifying the seismic fragility parameters that are calculated based on the site-specific median spectrum is described. This simple method was developed to incorporate the effects of the uniform hazard spectrum. The seismic fragility parameters of typical NPP components are modified using the uniform hazard spectrum. The modification factor is used to modify the original fragility parameters. An example uniform hazard spectrum is developed using the available seismic hazard data for the Korean nuclear power plant (NPP) site. This uniform hazard spectrum is used for the modification of fragility parameters.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1858-1868
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• 2021

Resisting an accidental impact of large commercial aircrafts is an important aspect of advanced nuclear power plant (NPP) design. Especially after the 9·11 event, some regulations were enacted, which required the design of NPPs should consider the accidental impact of large commercial aircrafts. Normal working of equipment is important for stopping reactor under an impact when an NPP is in operation. However, there is a lack of reliable analysis and research on the impact test of nuclear prototype equipment. Therefore, in order to study the response of the equipment under high acceleration impact, a centrifugal pump is selected as the research object to perform the impact test. A horizontal half-sinusoidal pulse wave was applied to the working pump. The test results show that the horizontal response of the motor and flange is greater compared to other parts, as well as the vertical response of the coupling. The stress response of the pump body support and motor support is high, hence these parts should be considered in the design of the pump. Finally, combined with the damage and stress evaluation results of the pump under different amplitudes, the ultimate impact acceleration that the pump can withstand is given.