Papers by Author: Young Hwan Choi

Abstract: A piping system including straight pipes, elbows and tee branches in a nuclear power plant is mostly subjected to severe loading conditions with high temperature and pressure. In particular, the wall-thinning of an elbow due to flow accelerated corrosion is one of safety issues in the nuclear industry. In this respect, it is necessary to investigate the limit loads of an elbow with a wall-thinned part for evaluating integrity. In this paper, three dimensional plastic limit analyses are performed to obtain limit loads of an elbow with different bend angles as well as defect geometries under internal pressure and in-plane/out-of-plane bending moment. The limit loads are also compared with the results from limit load solutions of an uninjured elbow based on the von Mises yield criteria. Finally, the effects of significant factors, bend angle and defect shape, are quantified to estimate the exact load carrying capacity of an elbow during operation.

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.

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.

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.

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.

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.

Abstract: Signals acquired from a Combo calibration standard tube used to calibrate for inspection and evaluation of motorized rotating pancake coil probe signals from steam generator tubes. So, Combo tube signals should be consistent and accurate since they have strong influence on evaluation procedure of signals. However, motorized rotating pancake coil probe signals are very easily affected by various factors so that they can distort amplitudes and phase angles which are quantitative terms for signal evaluation. To overcome this problem, we explored possibility of using numerical simulation as a practical calibration tool for the evaluation of real field signals. In this study, we investigated the characteristics of a motorized rotating pancake coil probe and a Combo tube. And then we used commercial software to produce a set of calibration signals and compared to the experiments. Using simulated Combo tube signals, we evaluated deliberated single circumferential indication defects, and these results were compared with experimental signal evaluation results.

Abstract: Major nuclear components have been designed by conservative codes to prevent unanticipated fatigue failure. However, more realistic and effective assessment is necessary in proof of continued operation beyond the design life. In the present paper, three-dimensional stress and fatigue evaluation is carried out for pressurizer employing complex full geometry itself instead of conventional discrete subcomponents. For this purpose, temperature and mechanical stress transfer Green’s functions are derived from finite element analyses and applied to critical locations of pressurizer. In accordance with comparison of resulting stresses obtained from the Green’s function and detailed finite element analysis, suitability of the specific Green’s function is investigated. Finally, prototype of fatigue life assessment results is provided along with relevant ongoing activities.

Abstract: Conventionally, shielded-metal arc welding (SMAW) process has been applied to join pipes of reactor coolant loop, which caused defects and lot of loss in time and cost due to excessive heat input in joining section. Recently, narrow-gap welding (NGW) process was introduced to overcome the disadvantages of SMAW. However, the application of NGW to nuclear power plant is not yet commonly used, because safety of NGW process is not fully proven. In the present paper, welded coupons are made of stainless steel. They are manufactured under different processes; general welding (GW), and repair welding after GW. Performed are various mechanical tests to investigate microstructure, tensile strength and so on. It is verified that the mechanical properties of stainless steel are slightly changed after repair welding process. It is also found from stress corrosion cracking tests that the failure time of repair welding is shorter than that of general welding.