Papers by Author: Chi Yong Park

Abstract: Inconel alloy such as alloy 600 and alloy 690 is widely used as the steam generator tube materials in the nuclear power plants. The impact fretting wear tests were performed to investigate wear mechanism between tube alloy and 409 stainless steel tube support plates in the simulated steam generator operating conditions, pressure of 15MPa, high temperature water of 290°C and low dissolved oxygen(<10 ppb). From investigation of wear test specimens by the SEM and EDS analysis, hammer imprint, which is known to be an actual damaged wear pattern, has been observed on the worn surface, and fretting wear mechanism was investigated. Wear progression of impact-fretting wear also has been examined. It was observed that titanium rich phase contributes to the formation of voids and cracks in sub-layer of fretting wear damage by impact fretting wear.

Abstract: The present work presents plastic limit load solutions for branch junctions under internal pressure and in-plane bending, based on detailed three-dimensional (3-D) FE limit analyses using elastic-perfectly plastic materials. The proposed solutions are valid for a wide range of branch junction geometries; ratios of the branch-to-run pipe radius and thickness from 0.0 to 1.0, and the mean radius-to-thickness ratio of the run pipe from 5.0 to 20.0.

Abstract: This paper provides closed-form plastic limit load solutions for elbows with local wall thinning under in-plane bending, via three-dimensional (3-D), small strain FE limit analyses using elastic-perfectly plastic materials. Wide ranges of elbow and thinning geometries are considered.

Abstract: Fretting wear generated by flow induced vibration is one of the important degradation mechanisms of steam generator tubes in the nuclear power plants. Understanding of tube wear characteristics is very important to keep the integrity of the steam generator tubes to secure the safety of the nuclear power plants. Experimental examination has been performed for the purpose of investigating the impact fretting. Test material is alloy 690 tube and 409 stainless steel tube supports. From the results of experiments, wear scar progression is investigated in the case of impact-fretting wear test of steam generator tubes under plant operating conditions such as pressure of 15MPa, high temperature of 290
C and low dissolved oxygen. Hammer imprint that is actual damaged wear pattern, has been observed on the worn surface. From investigation of wear scar pattern, wear mechanism was initially the delamination wear due to cracking the hard oxide film and finally transferred to the stable impact-fretting pattern.

Abstract: Dissimilar weld region located at the cooling tube of pump was damaged and failed. We performed a root cause analysis using the Scanning Electron Microscopy (SEM) and Energy Dispersive X-ray Spectroscopy for the fractured surface. Many internal defects were shown in the fracture section. Root cause analysis shows that the failure is due to welding at higher temperature over 1000
. This was confirmed by high-temperature cracks observed with SEM. In this study Failure scenario including initial crack generation, propagation of corrosion cracks and final failure, has been constructed and verified.

Abstract: The objective of this study is to develop a local failure criterion at wall-thinning defect of piping components. For this purpose, a series of tensile tests was performed using several types of specimens with different stress state under tension, including smooth bars, notched round bars and grooved plates. In addition, finite element (FE) simulations were performed for all tests, and its results were compared to the test results. From the comparisons, the equivalent stress and strain corresponding to maximum load and final failure of the notched round bar specimens were proposed as the local failure criterion which is a function of stress triaxiality at notched area. The criteria were verified by employing them to the estimation of maximum load and final failure of grooved plate specimen tests.

Abstract: In steam generators of nuclear power plants, flow-induced vibration (FIV) can lead to tube damage by fretting-wear occurred due to impact and sliding movement between the tubes and their supports. There have been many studies and test results on wear damage of steam generator tubes but they were not reflected the mechanical and chemical conditions accurately. KEPRI nuclear power laboratory developed a wear test system, which is able to control the motion of impact and sliding simultaneously in the pressurized high temperature water-chemistry conditions. Some wear tests were performed to verify the stable operation for the wear test. This wear test system with new concepts was described briefly, and some data for verifying its performance have been shown in the cases of the selected some test results. In the test, Alloy 690 was used for tube materials and 409 stainless steel for support plates. A little data deviation was obtained and stability of system operation was investigated.

Abstract: Fretting is the oscillatory motion with very small amplitudes, which usually occurs between two solid surfaces in contact. Fretting wear is the removal of material from contacting surfaces through fretting action. Fretting wear of steam generator tubes in nuclear power plant becomes a serious problem in recent years. The materials for the tubes usually are INCONEL 690 (I-690) and INCONEL 600 (I-600). In this paper, fretting wear tests for I-690 and I-600 were performed under various applied loads in water at room temperature. Results showed that the fretting wear loss of I-690 and I-600 tubes was largely influenced by stick-slip. The fretting wear mechanisms were the abrasive wear in slip regime and the delamination wear in stick regime. Also, I-690 had somewhat better wear resistance than I-600.

Abstract: In order to study the influence of stress ratio and WC grain size, the characteristics of fatigue crack growth were investigated in WC-Co cemented carbides with two different grain sizes of 3 and 6 µm. Fatigue crack growth tests were carried out over a wide range of fatigue crack growth rates covering the threshold stress intensity factor range DKth. It was found that crack growth rate da/dN against stress intensity factor range DK depended on stress ratio R. The crack growth rate plotted in terms of effective stress intensity factor range DKeff still exhibited the effect of microstructure. Fractographic examination revealed brittle fracture at R=0.1 and ductile fracture at R=0.5 in Co binder phase. The amount of Co phase transformation for stress ratio was closely related to fatigue crack growth characteristics.