Papers by Author: Young Hwan Choi

Paper TitlePage

Authors: Sung Jin Song, Joon Soo Park, Hak Joon Kim, Un Hak Seong, Suk Chull Kang, Young Hwan Choi
Abstract: In this study, the expanded multi-Gaussian beam model is adopted to develop a model to calculate the ultrasonic beam fields radiated from an ultrasonic phased array transducer. Combining this beam model with three other components including time delays, a far-field scattering model and a system efficiency factor, we develop a complete ultrasonic measurement model for predicting the phased array ultrasonic signals that can be captured from a flat-bottom hole in a steel specimen in a pulse-echo set-up using an array transducer mounted in a solid wedge. This paper describes the complete model developed with its key ingredients.
Authors: Sun Yeong Choi, Young Hwan Choi
Abstract: The current in-service inspection (ISI) strategy for the nuclear piping in many countries consists of both the code requirements such as ASME B & PV Code Sec. XI and the country-specific regulatory requirements, so called as the enhanced ISI. The enhanced ISI reflects the operating experience of piping failure, while the ASME Code Sec. XI requirement is based on random sampling for the inspection points. In this study, a new strategy for ISI of nuclear piping was proposed based on piping failure frequency. This strategy basically reflects the operating experience because the piping failure frequency is based on the piping failure database. The new concept of minimum inspection rate was also introduced in this new ISI strategy. As pilot study, the new ISI strategy was applied to the Class 1 piping system such as reactor coolant system and safety injection system of Ulchin Unit 5 which is the 1,000 MWe Korean Standard PWR. The results from the proposed new strategy were compared to those from the ASME Code Sec. XI. The results show that the new ISI strategy reasonably reflects the operating experience. The results also show that the concept of the minimum inspection rate can compensate the unbalance in the number of inspection points between the very large differences in the piping failure frequency.
Authors: Yoon Suk Chang, Seong In Moon, Young Jin Kim, Jin Ho Lee, Myung Ho Song, Young Hwan Choi
Abstract: In this paper, conservatisms of current plugging criteria on steam generator tubes are reviewed and six new failure prediction models for dual through-wall cracks are proposed. In order to determine the optimum ones among these local or global failure prediction models, a series of plastic collapse tests and corresponding finite element analyses are carried out with respect to two adjacent axial through-wall cracks in thin plates. Then, reaction force model, plastic zone contact model and COD (Crack Opening Displacement) base model were selected as the optimum ones for integrity assessment of steam generator tubes with dual cracks.
Authors: J.C. Kim, Sang Min Lee, Yoon Suk Chang, Jae Boong Choi, Young Jin Kim, Young Hwan Choi
Abstract: Steam generators working in nuclear power plants convert water into steam from heat produced in the reactor core and each of them contains from 3,000 to 16,000 tubes. Since these tubes constitute one of primary barriers under radioactive and high pressure condition, the integrity should be maintained carefully during the operation. The objective of this research is to introduce an integrity evaluation system for steam generator tubes as a substitute of well-trained engineers or experts. For this purpose, a couplet examination has been carried out on the complicated evaluation procedure and an efficient system named as STiES was developed employing three representative integrity evaluation methods: fracture mechanics analysis (crack driving force diagram and J-integral/Tearing modulus method) and limit load method. Exemplary analyses for steam generator tubes with various types of flaws showed good applicability of the proposed integrity evaluation system. So, it is anticipated that the system can be used for the calculation of reference pressure to decide either the continued operation or repair until next outage.
Authors: Jong Choon Kim, Jae Boong Choi, Young Jin Kim, Young Hwan Choi, Youn Won Park, Shinobu Yoshimura
Authors: J.C. Kim, Jae Boong Choi, Yoon Suk Chang, Young Jin Kim, Youn Won Park, Young Hwan Choi
Abstract: While the demand on electric power is consistently increasing, public concerns and regulations for the construction of new nuclear power plants are getting restrict, and also operating nuclear power plants are gradually ageing. For this reason, the interest on lifetime extension for operating nuclear power plants by applying lifetime management system is increasing. The 40-year design life concept was originally introduced on the basis of economic and safety considerations. In other words, it was not determined by technological evaluations. Also, the transient design data which were applied for fatigue damage evaluation were overly conservative in comparison with actual transient data. Therefore, the accumulation of fatigue damage may result in a big difference between the actual data and the design data. The lifetime of nuclear power plants is mostly dependent on the fatigue life of a reactor pressure vessel, and thus, the exact evaluation of fatigue life on a reactor pressure vessel is a crucial factor in determining the extension of operating life. The purpose of this paper is to introduce a real-time fatigue monitoring system for an operating reactor pressure vessel which can be used for the lifetime extension. In order to satisfy the objectives, a web-based transient acquisition system was developed, thereby, real-time thermal-hydraulic data were reserved for 18 operating reactor pressure vessels. A series of finite element analyses was carried out to obtain the stress data due to actual transient. The fatigue life evaluation has been performed based on the stress analysis results and, finally, a web-based fatigue life evaluation system was introduced by combining analysis results and on-line monitoring system. Comparison of the stress analysis results between operating transients and design transients showed a considerable amount of benefits in terms of fatigue life. Therefore, it is anticipated that the developed web-based system can be utilized as an efficient tool for fatigue life estimation of reactor pressure vessel.
Authors: Joo Young Yoo, Sung Jin Song, Chang Hwan Kim, Hee Jun Jung, Young Hwan Choi, Suk Chull Kang, Hyun Kyu Jung
Abstract: In the present study, the synthetic signals from the combo tube are simulated by using commercial electromagnetic numerical analysis software which has been developed based on a volume integral method. A comparison of the simulated signals to the experiments is made for the verification of accuracy, and then evaluation of five deliberated single circumferential indication signals is performed to explore a possibility of using a numerical simulation as a practical calibration tool. The good agreement between the evaluation results for two cases (calibration done by experiments and calibration made by simulation) demonstrates such a high possibility.
Authors: Sun Yeong Choi, Young Hwan Choi
Abstract: The purpose of this paper is to evaluate the piping failure frequency based on the piping failure events in Korean pressurized water reactors (PWRs) until the end of 2003. Two types of the piping failure frequencies including the piping damage frequency and the piping rupture frequency are considered in this study. The piping damage frequency for the failed piping system was estimated by using the piping population data such as the weld count or the base metal count. The piping rupture frequency related to the initiating event in a probabilistic safety assessment (PSA) was evaluated by using both the Bayesian approach (Method 1) and the conditional rupture probability approach (Method 2). In the Bayesian approach, two methods using Jeffreys noninformative prior (Method 1-1) and prior distributions based on the results in NUREG/CR-5750 (Method 1-2) were considered. Thirty piping failure events in ASME safety class pipings of Korean PWRs were identified and analyzed in this study. The results showed that the piping damage frequency for the events ranged from 5.42E-3/cr.yr to 2.77E-5/cr.yr. Three kinds of initiating events including the very small LOCA, the feedwater line break, and the flood are evaluated for Korean PWRs. The results for the piping rupture frequency in Korean PWRs were as follows: 1) The mean piping rupture frequency of the very small LOCA event ranged from 3.6E-3/cr.yr to 1.2E-2/cr.yr, the feedwater line break event from 3.6E-3/cr.yr to 2.5E-2/cr.yr, and the flood event from 7.8E-4/cr.yr to 3.6E-3/cr.yr. The mean piping rupture frequencies of the very small LOCA and feedwater line break events were higher than that of the flood event by one order of a magnitude. 2) Method 2 gave conservative results in the very small LOCA and feedwater line break events compared to Method 1-1 or Method 1-2, while Method 1-1 gave conservative results in the flood event. 3) The order of magnitudes in the mean piping rupture frequencies of the very small LOCA, the feedwater line break, and the flood in Korean PWRs were similar to those in the U.S. PWRs.
Authors: Jae Do Kwon, Seung Wan Woo, Young Hwan Choi
Abstract: A dissimilar weld zone exists between the pipe and nozzle in a primary reactor cooling system (RCS). Thermal aging is observed in cast stainless steel, CF8M used in a pipe as the RCS is exposed for a long period of time to a reactor operating temperature between 290 and 330°C. No effect is observed in low-alloy steel. SA508 cl.3 is used in a nozzle. The artificially accelerated aging specimens are prepared to maintain for a temperature of 430°C for 300, 1800, and 3600hrs, respectively. Then, various mechanical tests such as hardness, tension, impact test, are performed in virgin and aged specimens in order to determine the existence of dissimilar weld zones. The specimens for elastic-plastic fracture toughness tests are prepared for one type, where a notch is created in the heat affected zone of CF8M. From the experiments, it was found that J-integral values decrease as age increases.
Authors: J.C. Kim, M.Y. Ahn, Yoon Suk Chang, Jae Boong Choi, Young Jin Kim, Myung Jo Jhung, Young Hwan Choi
Abstract: In general, the fatigue life of major nuclear components has been evaluated based on design codes conservatively. However, sometimes, more exact fatigue life evaluation is required for continued operation beyond the endorsed life. The purpose of this paper is to carry out 3-D stress and fatigue analyses reflecting full geometry as well as actual operating data. The actual operating data acquired through a monitoring system were filtered and assessed. Then, temperature and stress transfer Green’s functions were developed and applied to critical locations of reactor pressure vessel. The finite element analyses results for representative design transients were verified through comparison to reference solution and showed that the conservatism of current 2-D evaluation. Therefore, it is anticipated that the proposed scheme adopting Green’s function and real operating histories can be utilized for remaining life time evaluation of major components.
Showing 1 to 10 of 23 Paper Titles