• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.870-875
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• 2023

In this paper, we analyzed the revised guidance on electromagnetic (EM) interference qualification of Regulatory Guide 1.180 (Revision 2), which is published by the U.S. Nuclear Regulatory Commission for electromagnetic compatibility qualification for nuclear safety, by comparing it with that of the previous version. We confirmed that the test methods and the acceptance criteria of both CE101 and CE102 tests for conductive emission and RE102 test for radiating emission are changed in the recently revised Regulatory Guide 1.180 (Revision 2). Furthermore, we found that the revised Regulatory Guide 1.180 provides flexibility in the use of alternative methods for EM interference (EMI) qualification, in that a mix of the various base-standards is technically allowed. In addition, the primary revision of the updated Regulatory. Guide 1.180 is that MIL-STD-461G is to be adopted as the latest base-standard, instead of MIL-STD-461E. To evaluate the influence on EMI qualification for nuclear safety due to the endorsement of MIL-STD-461G, we thoroughly analyzed the modifications in the acceptance criteria and test methods for EMI qualification, and then validated the analyzed effect on the EMI qualification, which is caused by the revision of MIL-STD-461, by performing electromagnetic simulation for equipment under RE102 test.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.15-22
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• 2011

The chemistry factor and RTPTS margin for domestic reactor pressure vessel materials were analyzed by using the surveillance data which have been obtained from 8 nuclear power plants in Korea. The surveillance data have been used to assess the integrity of the pressure vessel under the pressurized thermal shock (PTS) event. The chemistry factor, which is determined by the Cu and Ni contents of vessel materials, is considered a proper tool to assess the $RT_{PTS}$. The chemistry factors, which were obtained from the surveillance data of domestic reactor pressure vessels, were investigated and compared with those of Regulatory Guide 1.99 in this study. Regressions for ${\Delta}RT_{NDT}$ were performed to expect the chemistry factor as a function of Cu and Ni, and to estimate $RT_{PTS}$ margin. The margin analysis was performed by comparing the regression graphs and standard deviations with those of Regulatory Guide 1.99. The standard deviations calculated by using the domestic surveillance data for base metal and welds are almost same as the standard deviations which are suggested on Regulatory Guide 1.99, Rev.2.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2017.11a
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• pp.63-64
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• 2017

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.

• Journal of the Korean Society of Safety
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• v.37 no.5
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• pp.74-79
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• 2022

This study analyzes the contents of the second revised Regulatory Guide 1.180 (Revision 2) for electromagnetic compatibility qualification published by the U. S. Nuclear Regulatory Commission by comparing them with those of the previous version. The methods and acceptance criteria of both CE101 and CE102 tests pertaining to conductive emission and RE102 test for radiation emission are observed to have been modified in Revision 2. Furthermore, the revised guide is found to afford flexibility in using alternative methods for electromagnetic interference qualification by allowing the combination of different technical base standards.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2012.10a
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• pp.306-311
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• 2012

A reactor vessel internals comprehensive vibration assessment program (RVI CVAP) of an advanced power reactor 1400 (APR1400) is being performed as a non-prototype category-2 type of reactor based on the US Nuclear Regulatory Commission Regulatory Guide (NRC RG) 1.20. The aim of this paper is to present the results of structural response analysis and measuring locations of a upper guide structure (UGS) assembly of the APR1400 reactor. The analysis results of the UGS assembly results show that meet the specified integrity levels of the design acceptance criteria. Also, the measuring locations are set by the analysis results of the UGS assembly and selection criteria of measuring locations prior to this study. These analysis results and measuring locations will be used as fundamental materials to design a measurement system for the APR1400 RVI CVAP.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.963-970
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• 2018

The U.S. Nuclear Regulatory Commission, Regulatory Guide 1.20 needs to perform a comprehensive vibration assessment program for reactor internals during preoperational and startup testing for nuclear power plants and extended power uprate. Although the measurement program is one of the core programs, it is rarely carried out except for a first-of-a-kind or a unique design. This article describes measurement results of vibration and stress for the comprehensive vibration assessment program for an APR-1400 reactor internals. The measurement was performed at an upper guide structure during the pre-core hot functional test of Shin Kori unit 4 reactor internals because the Shin Kori unit 3 and 4 are the first construction project for the APR-1400, and the upper guide structure assembly was to design change compared with the valid prototype. We confirmed that all measured results are within the test acceptance criteria. It means that the structural integrity of the APR-1400 reactor internals was secured for the flow-induced vibration.

• Tunnel and Underground Space
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• v.24 no.5
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• pp.354-365
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• 2014

The horizontal response spectra using the observed ground motions from 15 Fukuoka earthquake series, including main earthquake (2005/03/20; Mw=6.5), were analysed and then were compared to both the seismic design response spectra (Regulatory Guide 1.60), applied to the domestic nuclear power plants, and the Korean Standard Design Response Spectrum for general structures and buildings (1997). 178 horizontal ground motions, without considering soil types, were used for normalization with respect to the peak acceleration value of each ground motion. The results showed that response spectrum have strong dependency on epicentral distance. The results also showed that the horizontal response spectra revealed higher values for frequency bands at two frequency bands (about 8 - 10 and 16- 20 Hz) than Regulatory Guide 1.60. The results were also compared to the Korean Standard Building Design Spectrum for the 3 different soil types and showed that thehorizontal response spectra revealed higher values for almost all period bands than the Korean Standard Response Spectrum (500 yrs; Return Period; Seismic province 1; SE soil conditions). Through the qualitative improvements and quantitative enhancement of the observed ground motions, the diversity of the observed ground motions should be considered more significantly to improve the certainty of response spectrum.

• Proceedings of the KIEE Conference
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• 2007.11b
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• pp.66-68
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• 2007

The regulatory guide, RG-1.204 and its underlying or confirmatory research, NUREG/CR-6866 were studied for Korean application to LPGS of NPPs. However they excluded the application to that of NPPs. So US-NRC approved selectively industrial standards to that of NPPs on Nov 2005. It is necessary to understand the basis of regulatory technology related regulatory positions on LPGS and important to implement the guidance on LPGS as a resonable standard. The paper is examined what and how state of the art of relevant technology applied to the LPGS as well as the trip-out events related to electrical system were involved with LPGS. We reviewed the relevant standards applicable to Korean NPPs. Following are concluded to recommend. (1) IEEE 510-1050 is recommended as a guide for I&C grounding against EMI and lighting transients (2) IEEE Std-665, 510-666, Std-C62.23 for electrical grounding against voltage surges and lighting transients (3) Inspection should be thoroughly be implemented a frequency of 3-5 year period according to NFPA780 or KSC-IEC 61024.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1791-1798
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• 2019

Instrumentation and Control (I&C) systems of nuclear power plants (NPPs) have been continuously digitalized. These systems have a critical role in the operation of nuclear facilities by functioning as the brain of NPPs. In recent years, as cyber security threats to NPP systems have increased, regulatory and policy-related organizations around the world, including the International Atomic Energy Agency (IAEA), Nuclear Regulatory Commission (NRC) and Korea Institute of Nuclear Nonproliferation and Control (KINAC), have emphasized the importance of nuclear cyber security by publishing cyber security guidelines and recommending cyber security requirements for NPP facilities. As described in NRC Regulatory Guide (Reg) 5.71 and KINAC RS015, challenge response authentication should be applied to the critical digital I&C system of NPPs to satisfy the cyber security requirements. There have been no cases in which the most robust response authentication technology like challenge response has been developed and applied to nuclear I&C systems. This paper presents a challenge response authentication mechanism for a Programmable Logic Controller (PLC) system used as a control system in the safety system of the Advanced Power Reactor (APR) 1400 NPP.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.23 no.1
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• pp.49-55
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• 2013

A reactor vessel internals comprehensive vibration assessment program(RVI CVAP) of an advanced power reactor 1400(APR1400) is being performed as a non-prototype category-2 type of reactor based on the US nuclear regulatory commission regulatory guide(NRC RG) 1.20. The aim of this paper is to present the results of structural response analysis and measuring locations of a upper guide structure(UGS) assembly of the APR1400 reactor. The analysis results of the UGS assembly show that the specified integrity levels meet the design acceptance criteria. Also, the measuring locations are determined by the analysis results of the UGS assembly and selection criteria of previous study. These analysis results and measuring locations will be used as a guide to design a measurement system for the APR1400 RVI CVAP.