• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.1
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• pp.43-49
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• 2010

The piping failure probability of the reactor coolant system in Kori unit 1 was evaluated considering stress corrosion cracking. The P-PIE program (Probabilistic Piping Integrity Evaluation Program) developed in this study was used in the analysis. The effect of some variables such as oxygen concentration during start up and steady state operation, and operating temperature, which are related with stress corrosion cracking, on the piping failure probabilities was investigated. The effects of leak detection capability, the size of big leak, piping loops, and reactor types on the piping failure probability were also investigated. The results show that (1) LOCA (loss of coolant accident) probability of Kori unit 1 is extremely low, (2) leak probability is sensitive to oxygen concentration during steady state operation and operating temperature, while not sensitive to the oxygen concentration during start up, and (3) the piping thickness and operating temperature play important roles in the leak probabilities of the cold leg in 4 reactor types having same inner diameter.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.1-7
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• 2013

In around 2000, the U.S. NPPs have developed the various advanced engineering processes based on the INPO AP-913(Equipment Reliability Process Description) and showed the high performance in availability. With these benchmarking cases, the Korean NPPs have introduced the advanced engineering technology since 2005. The first step of the advanced engineering is to analyze and determine component importance for all components of a plant. This process is called Functional Importance Determination(FID). These results are basically utilized to determine the priority with limited resources in various areas. However, because the consistency of FID results is insufficient despite applying the same criteria in the existing operating NPPs, the degree of application is low. Therefore, this paper presents the improved methodology for FID interfacing system functions of Maintenance Rule Program and results of Single Point Vulnerability(SPV). This improved methodology is expected to contribute to enhance the reliability of FID data.

• The Journal of the Korea Contents Association
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• v.14 no.3
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• pp.463-469
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• 2014

Safety related devices unconstrained temporally in the process of operation of nuclear power plants could be damaged by the sliding during seismic activity. In this study sliding response of unconstrained objects to the base excitations is studied experimentally and analytically. In experiments static and dynamic tests to determine the coefficient of friction and the shaking table experiments to verify the sliding response of the analytic results were conducted. Numerical solutions by solving the nonlinear differential equations of motion governing sliding were found by the computer program using the step by step acceleration method. The exact solutions of the sliding response to the simple forms of base excitations were found to verify the computer program developed in this study. Relative displacement envelopes were suggested as a colliding criteria of the unconstrained objects.

• Journal of the Korean Society for Nondestructive Testing
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• v.15 no.1
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• pp.310-317
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• 1995

Crack detection technique by ultrasonics in structures and components made of austenitic stainless steel often loses its reliability due to the material characteristics during inservice inspection of nuclear power plants, especially in the area of detection and sizing in centrifugally cast stainless steel pipings. In order to understand and overcome this problem, computer program for tracing the ultrasonic rays within material has been developed to simulate the process of defect detection within weld. The program simulates through transmission and reflection technique in crack detection of austenitic stainless steel as well as ultrasonic beam propagation through multiple media including stainless steel cladding interface.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.10a
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• pp.99-101
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• 2014

A Comprehensive Vibration Assessment Program (CVAP) for steam generator internals (SGI) of Advanced Power Reactor 1400 (APR1400) is being performed in accordance with the United States Nuclear Regulatory Commission (U.S. NRC) Regulatory Guide 1.20 (RG 1.20) revision 3. This paper studies the vibration characteristics of moisture separator assembly as part of the vibration and stress analysis program for APR1400 SGI CVAP. The natural frequencies, mode shapes, and structural behavior of moisture separator assembly were investigated through modal analysis using finite element method and experimental measurement. Since the moisture separator consists of several items with complicated shape, an idealized shell model was used in the finite element analysis. Group of local modes caused by moisture separators and significant modes of shroud and separator support plate were identified. The results of this paper are to be utilized in the structural response analysis of moisture separator assembly.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.34-39
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• 2008

Steam Generator(SG) tubes are classified as one of the key components in nuclear power plants, and they should be periodically examined by the intensified management program for the assurance and diagnosis of their structural integrity. In this study, we use the optimum analysis technique to draw the detection and categorization of bowing(BOW) signals; abnormal tube-to-tube proximity in the SG upper bundle free span area. The locations in which BOW signals are detected likely have latent degradation of ODSCC(Outer Diameter Stress Corrosion Cracking). For the sake of timely and correct detection of BOW signals and diagnosis of ODSCC, we carried out the experimental demonstrations using a reduced mock-up. And we validated the MRPC(Motorized Rotating Pancake Coil) analysis technique is better than the bobbin. Hence, it comes to conclusion that the optimum analysis technique can be a good alternative for the reliable SG tube examination.

• Proceedings of the Korea Concrete Institute Conference
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• 1998.10b
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• pp.686-693
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• 1998

Prestressed Concrete Cylinder Pipe(PCCP) is used in water transmission pipeline of the Great Man-made River Project(GMR Project). In domestic area, PCCP is used for water cooling systems of Uljin and Youngkwang nuclear power plants. In abroad, especially in the United States and Canada PCCP supplies virtually every metropolitan area with raw and treated water. Compared with other pipe types, PCCP manufacturing cost is dear. But total cost can be considered as economical due to low installation and maintenance cost. Previously, the designs of PCCP were generally determined from one of two appendices in American Water Works Association(AWWA)standard C301 which provided two design methods-cubic parabola design method and stress analysis design method. In 1992, the design procedure for PCCP expanded from two alternatives to the most huge and complex AWWA standard C304. Because C304is so large, it takes too much time for the engineer to read and understand the design concepts and procedures. In this paper, the AWWA C304 design procedures are segmented into simple, understandable sections and concepts and explained. Each section or concepts is compared to the previous design procedure to highlight the revisions and reasons for them. Also the PCCP design program was developed and the design program results are compared with the calculations of the GMR project design consultant.

• Proceedings of the KSR Conference
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• 2008.11b
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• pp.1487-1493
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• 2008

Railway vehicle is very important to implement the effective maintenance in proper to prevent any failure during operation period. Many railway authorities are making efforts to maintain the railway vehicle through scientific and systematic procedure. To achieve this, Reliability Centered Maintenance(RCM) is partially applied. The efficiency of RCM has proven and its terminology was familiar with nuclear power, military and chemical plant etc. since the commercial aircraft's industries has introduced the maintenance program based on the target of reliability. The application of RCM on railway vehicle can be utilized with systematic analysis method to select the best effective maintenance period and action to prevent the failures by selecting the equipment affecting the its safety and reliability. This paper is presented that the procedure of adequate and effective maintenance for railway vehicle by comparing among the related standards in example IEC60300-3,11, MIL-STD-2173, and technical documents or papers. In accordance with above result, RCM procedure is proposed to apply effectively for maintenance of railway vehicle. That is, (1) Analysis of data and Calculation of criticality per equipment (2) Selection of equipment to analyze (3) Analysis of failure mode and effect (4) Evaluation of maintenance method and period (5) Optimization of maintenance program through renewal of maintenance method and period.

• Korean Journal of Hazardous Materials
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• v.3 no.1
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• pp.28-36
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• 2015

In early 2011, the Fukushima nuclear power plant had greater damage due to earthquake in Japan, and the awareness of safety has increased. In particular, special response systems should be required to handle disaster situations in plant sites which are likely to occur for large disasters. In this study, a program is designed to set up optimum escape routes, by a smart phone application, when a disaster situation occurs. This program could get information of the cumulative damage from sensors and display the escape route of the smallest damage in real-time on the screen. Utilizing our application in real-time evacuation has advantage in reducing cumulative damage. The optimal evacuation route, focusing on horizontal path, is calculated based on getting the data of fire, detected radioactivity and hazardous gas. Thus, using our application provides information of optimal evacuation to people who even can not hear sensor alarms or do not know geography, without requiring additional costs except fixed sensors or server network deployment cost. As a result, being informed of real-time escape route, the user could behave rapidly with suitable response to individual situation resulting in improved evacuation than simply reacting to existing warning alarms.

• Computational Structural Engineering
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• v.5 no.4
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• pp.93-102
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• 1992

The design of reactor internals requires the accurate vibration characteristics of each component for subsequent dynamic structural response analyses. For Korean standard nuclear power plant some modifications on the Control Element Assembly shroud from the reference design have been made, Since the shroud is complex in geometry having an array of vertical round tubes and webs in a square grid pattern, and being tied down by preloaded tie rods into position, it is planned to perform a vibration measurement program consisting of both experimental and analytical modal studies upon that component. The shroud modal testing was performed on the low frequency global survey to measure the first several modes. The analysis using the finite element model was also performed for the as-tested conditions. The natural frequencies and mode shapes from both test and analysis have been acquired and compared to be in good agreement. It is concluded that finite element model generated is good enough to be used in the design for the dynamic response analysis under various loading conditions.