• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2013.04a
• /
• pp.589-595
• /
• 2013

A reactor vessel internals comprehensive vibration assessment program(RVI CVAP) of an advanced power reactor 1400(APR1400) is being verified on the integrity of RVI for the design life of the plant by performing the non-prototype category-2 type on the US Nuclear Regulatory Commission Guide(NRC RG) 1.20, for which consists of a vibration and stress analysis program, a limited vibration measurement program, an inspection program, and the correlation of these programs. The aim of this paper is to describe the plan for the vibration measurement, test and acceptance criteria portion, and documentation and results of the APR1400 RVI CVAP. We will conduct the limited vibration measurement program of the APR1400 RVI CVAP according to the measurement plan and the vibration measurement testing in this paper.

• Journal of Electrical Engineering and Technology
• /
• v.9 no.4
• /
• pp.1369-1374
• /
• 2014

Most accidents of medium-voltage cables installed in nuclear power plants result from the initial defect of internal insulators or the initial failure due to poor construction. However, as the service years of plants increase, the possibility of cable accidents is also rapidly increases. This is primarily caused by electric, mechanical, thermal, and radiation stresses. Recently, much attention is paid to the study of cable diagnoses. To date, partial discharge and Tan${\delta}$ measurements are known as reliable methods to diagnose the aging of medium-voltage cables. High frequency partial discharge measurement techniques have been widely used to diagnose cables in transmission and distribution systems. However, the on-line high frequency partial discharge technique has not been used in the nuclear power plants because of the plant shutdown risk, degraded measurement sensitivity, and application problems. In this paper, the partial discharge measurement with a portable device was tried to evaluate the integrity of the 4.16kV and 13.8kV cable lines. The test results show that the high detection sensitivity can be achieved by the high frequency partial discharge technique. The present technique is highly attractive to diagnose medium voltage cables in nuclear power plants.

• The Transactions of The Korean Institute of Electrical Engineers
• /
• v.64 no.2
• /
• pp.254-259
• /
• 2015

On January 30, 2012 an auxiliary component of Byron Unit 2 was tripped on bus under voltage. The cause of the event was the failure of the C-phase insulator track for the Unit 2 station auxiliary transformer(SAT) revenue metering transformer. In addition to this event, other events have occurred at other plants resulting in an open-phase condition.[1] Therefore, Nuclear Regulatory Commission(NRC) has requested that not only nuclear power plant(NPP) operating company but also its Design Certification(DC) applicant have to prepare open-phase detecting system in their operating plants and design document. In this paper, various open-phase conditions are simulated in NPP using Electromagnetic Transient Program(EMTP) and Atpdraw, and open-phase condition detecting system is proposed for Main Transformer(MT), Unit Auxiliary Transformer(UAT) and SAT connected power line in NPP.

• Journal of the Earthquake Engineering Society of Korea
• /
• v.16 no.3
• /
• pp.35-42
• /
• 2012

We propose a probabilistic method to evaluate the uniform hazard spectra (UHS) of the soil of nuclear power plant(NPP) sites corresponding to that of a bedrock site. To do this, amplification factors on the surface of soil sites were estimated through site response analysis while considering the uncertainty in the earthquake ground motion and soil deposit characteristics. The amplification factors were calculated by regression analysis with spectral acceleration because these two factors are mostly correlated. The proposed method was applied to the evaluation of UHS for the KNGR (Korean Next Generation Reactor) and the APR1400 (Advanced Power Reactor 1400) nuclear power plant sites of B1, B4, C1 and C3. The most dominant frequency range with respect to the annual frequency of earthquakes was evaluated from the UHS analysis. It can be expected that the proposed method will improve the results of integrated risk assessments of NPPs rationally. We expect also that the proposed method will be applied to the evaluation of the UHS and of many other kinds of soil sites.

• Nuclear Engineering and Technology
• /
• v.43 no.1
• /
• pp.89-98
• /
• 2011

The construction technology for reactor vessel internals (RVI) modularization is one of the most important factors to be considered in reducing the construction period of nuclear power plants. For RVI modularization, gaps between the reactor vessel (RV) core-stabilizing lug and the core support barrel (CSB) snubber lug must be measured using a remote method from outside the RV. In order to measure RVI gaps remotely at nuclear power plant construction sites, certain core technologies must be developed and verified. These include a remote measurement system to measure the gaps between the RV core-stabilizing lug and the CSB snubber lug, an RVI mockup to perform the gap measurement tests, and a new procedure and schedule for RVI installation. A remote measurement system was developed previously, and a gap measurement test was completed successfully using the RVI mockup. We also developed a new procedure and schedule for RVI installation. This paper presents the new and improved installation procedure and schedule for RVI modularization. These are expected to become core technologies that will allow us to shorten the construction period by a minimum of two months compared to the existing installation procedure and schedule.

• Proceedings of the Society of Korea Industrial and System Engineering Conference
• /
• 2002.05a
• /
• pp.25-30
• /
• 2002

The problem of qualitative reliability system is very important issue in the digitalized nuclear power plant, because the failure of its system brings about extravagant economic loss, extensive environment destruction, and fatal damage of human. Therefore this study is to develop the reliability evaluation model through the normalized scoring model by the quantitative and qualitative factors considering the advanced safety factors In the Advanced Pressurized water Reactor 1400MWe(APR 1400) under uncertainty Especially, the qualitative factors considering the information and human factors for the systematic and rational justification have been closely analyzed. The reliability evaluation model can be simply applied in real fields in order to minimize the industrial accident and human error in the digitalized nuclear power plant.

• IE interfaces
• /
• v.22 no.2
• /
• pp.104-115
• /
• 2009

The objective of the study is to evaluate the effectiveness of Crew Resource Management (CRM) training program for operators in the Main Control Room (MCR) simulator of APR-1400 Nuclear Power Plant. The experiments were conducted for two different crews of operators performing six different emergency operating scenarios during four-week period. Each crew consisted of the five operators: senior reactor operator, safety technical advisor, reactor operator, turbine operator, and electric operator. All crews (Crew A and B) participated in the training program for the technical knowledge and skills which were required to operate the simulator of the MCR during the first week. To verify the effectiveness of the CRM training program; however, only Crew A was selected to attend the CRM training after the technical knowledge and skills training. The results of the experiments showed that the CRM training program improved the individual attitudes of Crew A significantly. Team skills of Crew A were found to be significantly better than those of Crew B. The CRM training did not have positive effects on enhancing the individual performance of Crew A; however, as compared to that of Crew B. Implication of these findings was discussed further in detail.

• Journal of the Korean Society of Systems Engineering
• /
• v.14 no.2
• /
• pp.58-66
• /
• 2018

This paper represents a systematic approach aimed at improving the performance of the proportional integral (PI) controller for the Advanced Power Reactor (APR) 1400 Feedwater Control System (FWCS). When the performance of the PI controller offers superior control and enhanced robustness, the steam generator (SG) level is properly controlled. This leads to the safe operation and increased the availability of the nuclear power plant. In this paper, a systems engineering approach is used in order to design a novel PI controller for the FWCS. In the reverse engineering stage, the existing FWCS configuration, especially the characteristics of the feedwater controller as well as the feedwater flow path to each SG from the FWCS, were reviewed and analysed. The overall block diagram of the FWCS and the SG was also developed in the reverse engineering process. In the re-engineering stage, the actual design of the feedwater PI controller was carried out using a genetic algorithm (GA). Lastly, in the validation and verification phase, the existing PI controller and the PI controller designed using GA method were simulated in Simulink/Matlab. From the simulation results, the GA-PI controller was found to exhibit greater stability than the current controller of the FWCS.

• Journal of the Korean Society of Systems Engineering
• /
• v.13 no.2
• /
• pp.9-17
• /
• 2017

Concept development of a real-time Field Programmable Gate Array (FPGA)-based switched Ethernet data communication network for the Man-Machine Interface System (MMIS) is presented in this paper. The proposed design discussed in this research is based on the systems engineering (SE) approach. The design methodology is effectively developed by defining the concept development stage of the life-cycle model consisting of three successive phases, which are developed and discussed: needs analysis; concept exploration; and concept definition. This life-cycle model is used to develop an FPGA-based time-triggered Ethernet (TTE) switched data communication network for the non-safety division of MMIS system to provide real-time data transfer from the safety control systems to the non-safety division of MMIS and between the non-safety systems including control, monitoring, and information display systems. The original IEEE standard 802.3 Ethernet networks were not typically designed or implemented for providing real-time data transmission, however implementing a network that provides both real-time and on-demand data transmission is achievable using the real-time Ethernet technology. To develop the design effectively, context diagrams are implied. Conformance to the stakeholders needs, system requirements, and relevant codes and standards together with utilizing the TTE technology are used to analyze, synthesize, and develop the MMIS non-safety data communication network of the APR1400 nuclear power plant.

• Journal of the Korean Society of Systems Engineering
• /
• v.12 no.2
• /
• pp.91-99
• /
• 2016

The safety of nuclear power plants has received much attention; this safety largely depends on the continuous availability of electrical energy source during all modes of nuclear power plant operation. A station blackout (SBO) describes the loss of the off-site electric power, the failure of the emergency diesel generators, and the unavailability of the alternate AC (AAC) power. Consequently, all systems that are AC powered such as the safety injection, shutdown cooling, component cooling water, and essential service water systems are unavailable. The aim of this study is to investigate the deficiencies of the existing alternatives for coping with an extended SBO for APR1400 design. The method is analyzing the existing deficiencies and proposing an optimal solution for the NPP design during the extended SBO. This study, established a new passive system, called passive decay heat removal system (PDHRS), using systems engineering approach.