• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2011.10a
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• pp.221-226
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• 2011

The structural integrity of APR1400 reactor vessel internals has been being assessed referring the US Nuclear Regulatory Commission regulatory guide 1.20 comprehensive vibration assessment program. The program is composed of a vibration and stress analysis, a limited vibration measurement, and an inspection. This paper covers the vibration and stress analysis on the reactor vessel internals due to the pulsation of reactor coolant pump. 3-dimensional models to calculate the hydraulic loads and structural responses were built and the pressure distributions and the structural responses were predicted using ANSYS. The peak stress of the reactor vessel internals is much lower than the acceptance limit.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.10a
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• pp.102-105
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• 2014

In accordance with U.S. Nuclear Regulatory Commission regulatory guide(NRC RG) 1.20(Rev.3), we are writing a comprehensive vibration assessment program(CVAP) design control document(DCD) and a technical report for U.S. NRC design certification(DC) of an Advanced Power Reactor 1400(APR1400) nuclear power plant(NPP). CVAP of an APR1400 NPP for U.S. NRC DC is classified as a non-prototype category 1 type. Therefore, CVAP DCD of reactor vessel internals(RVI) and steam generator internals(SGI) consist of analysis and full inspection program. However, piping system of primary and secondary system will be described as measurement program.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2018.05a
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• pp.318-319
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• 2018

Protective coatings at nuclear power plants should be designed to withstand exposure to ambient conditions during normal operation or design-basis accidents. However, there was a change in the perception of the protective coating to the revision of the Regulatory Guidelines by the NRC in July 2000. In other words, maintenance guidelines have been strengthened in order to minimize the clogging of the cooling water system due to the substances in the containment building. Therefore, KHNP, the contractor and operator of the nuclear power plant, plans to develop the coating system for nuclear power plants in accordance with the regulation, and plans to develop its own coating expert.

• International Journal of Safety
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• v.10 no.2
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• pp.1-5
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• 2011

The structural integrity assessment of APR1400 (Advanced Power Reactor 1400) reactor vessel internals has been being performed referring the US Nuclear Regulatory Commission regulatory guide 1.20 comprehensive vibration assessment program prior to commercial operation. The program is composed of a hydraulic and structural analysis, a vibration measurement, and an inspection. This paper describes the hydraulic and structural analysis on the reactor vessel internals due to hydraulic loads caused by the turbulence of reactor coolant. Three-dimensional models were built for the hydraulic and structural analysis and then hydraulic loads and structural responses were predicted for five analysis cases with CFX and ANSYS respectively. The structural responses show that the APR1400 reactor vessel internals have sufficient structural integrity in comparison with the acceptance criteria.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.309-316
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• 2024

The U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.20 provides guidance on the comprehensive vibration assessment program (CVAP) to be performed on reactor internals during preoperational and startup tests. The purpose of the program is to identify loads that could cause vibration in the reactor internals and to ensure that these vibrations do not affect their structural integrity. The structural vibrational analysis program involves creating finite element analysis models of the reactor internals and calculating their structural responses when subjected to vibration loads. The appropriateness of the structural analysis methodology must be demonstrated through benchmarks or any other reasonable means. Although existing structural analysis methodologies have been proven to be appropriate and are widely used, this paper presents the development of an improved new structural analysis methodology for APR1400 reactor internals using scaled model tests.

• Proceedings of the KIEE Conference
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• 2008.10b
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• pp.209-210
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• 2008

In the case of unauthorized individuals, systems and entities or process threatening the instrumentation and control systems of nuclear facilities using the intrinsic vulnerabilities of digital based technologies, those systems may lose their own required functions. The loss of required functions of the critical systems of nuclear facilities may seriously affect the safety of nuclear facilities. Consequently, digital instrumentation and control systems, which perform functions important to safety, should be designed and operated to respond to cyber threats capitalizing on the vulnerabilities of digital based technologies. To make it possible, the developers and licensees of nuclear facilities should perform appropriate cyber security program throughout the whole life cycle of digital instrumentation and control systems. Under the goal of securing the safety of nuclear facilities, this paper presents the KINS' regulatory position on cyber security program to remove the cyber threats that exploit the vulnerabilities of digital instrumentation and control systems and to mitigate the effect of such threats. Presented regulatory position includes establishing the cyber security policy and plan, analyzing and classifying the cyber threats and cyber security assessment of digital instrumentation and control systems.

• Asian Pacific Journal of Cancer Prevention
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• v.16 no.9
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• pp.3677-3683
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• 2015

Abnormally expressed microRNAs (miRNAs) and genes have been found to play key roles in esophageal squamous cell carcinoma (ESCC), but little is known about the underlying mechanisms. The aim of this paper was to assess inter-relationships and the regulatory mechanisms of ESCC through a network-based approach. We built three regulatory networks: an abnormally expressed network, a related network and a global network. Unlike previous examples, containing information only on genes or miRNAs, the prime focus was on relationships. It is worth noting that abnormally expressed network emerged as a fault map of ESCC. Theoretically, ESCC might be treated and prevented by correcting the included errors. In addition, the predicted transcription factors (TFs) obtained by the P-match method also warrant further study. Our results may further guide gene therapy researchers in the study of ESCC.

• Journal of Korea Society of Digital Industry and Information Management
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• v.16 no.1
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• pp.99-109
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• 2020

In the past, conservative concepts have been applied in terms of the characteristic of nuclear power plants(NPPs), resulting in analog-based equipment and closed networks. However, as digital technology has recently been applied to the design, digital-based facilities and communication networks have been used in nuclear power plants, increasing the risk of cybersecurity than using analog-based facilities. Nuclear power plant facilities are divided into a safety system and a non-safety system. It is essential to identify the difference and cope with cybersecurity threats to the safety system according to its characteristics. In this paper, we examine the cybersecurity regulatory guidelines for safety systems in nuclear power plant facilities. Also, we analyze cybersecurity threats to a programmable logic controller of the safety system and suggest cybersecurity requirements be applied to it to respond to the threats. By implementing security functions suitable for the programmable logic controller according to the suggested cybersecurity requirements, regulatory guidelines can be satisfied, and security functions can be extended according to other system requirements. Also, it can effectively cope with cybersecurity attacks that may occur during the operation of nuclear power plants.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.517-524
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• 2017

Cyber security is an important issue in the field of nuclear engineering because nuclear facilities use digital equipment and digital systems that can lead to serious hazards in the event of an accident. Regulatory agencies worldwide have announced guidelines for cyber security related to nuclear issues, including U.S. NRC Regulatory Guide 5.71. It is important to evaluate cyber security risk in accordance with these regulatory guides. In this study, we propose a cyber security risk evaluation model for nuclear instrumentation and control systems using a Bayesian network and event trees. As it is difficult to perform penetration tests on the systems, the evaluation model can inform research on cyber threats to cyber security systems for nuclear facilities through the use of prior and posterior information and backpropagation calculations. Furthermore, we suggest a methodology for the application of analytical results from the Bayesian network model to an event tree model, which is a probabilistic safety assessment method. The proposed method will provide insight into safety and cyber security risks.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.1109-1112
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• 2017

The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.