Papers by Author: Tae Eun Jin

Abstract: This study performed tensile test using small-size flat specimen and ball indentation test at room temperature to characterize the local tensile properties of bi-metallic weld joints. The weld specimens used were fabricated by joining between SA508 Gr.3 ferritic steel and Type 316 stainless steel with Alloy 82 buttering on the ferritic steel side and Alloy 82/182 weld metal. The test results showed that yield stress (YS) of weld metal was slightly higher than that of Type 316 and smaller than that of SA508 Gr.3, and ultimate tensile stress (UTS) of weld metal was similar as those of Type 316 and SA508 Gr.3 base metals. Also, the values of YS and UTS of buttering layer (Alloy 82) were nearly same as those of weld metal. Heat-affected-zones (HAZs) showed higher YS and UTS values compared to their base metals. Especially, the strengths of SA508 Gr.3 were significantly higher than those of surrounding materials. Also, it was known that the ball indentation test reasonably measured the local YS and UTS of bi-metallic weld joints.

Abstract: Steam generator in a nuclear power plant is huge heat exchanger that transfers heat from reactor to make steam to drive turbine-generator. Failure of the steam generator tubes can result in the release of fission products to the secondary side. Therefore, accurate integrity assessment of the cracked steam generator tubes is of great importance for maintaining the safety of the nuclear power plant. This paper provides limit loads for circumferential through-wall cracks in steam generator tubes under combined internal pressure and bending loads. Such limit loads are developed on the basis of three dimensional finite element analyses assuming elastic-perfectly plastic material behavior. As for the crack location, both the top of the tubesheet and U-bend regions are considered. The analysis results can be directly applied to the practical integrity assessment of cracked steam generator tubes, because the comparison between experimental data and FE results shows a very good agreement.

Abstract: The present work proposes a method for elastic-plastic fracture mechanics analysis of the circumferential through-wall crack in between elbows and attached straight pipes, subject to in-plane bending. Based on small strain finite element limit analyses, closed-form limit load solutions are given first. Then applicability of the reference stress based method to approximately estimate J is proposed. One interesting finding is that a popular approach to assume that the crack locates in the straight pipe could lead to significantly different assessment results.

Abstract: Kori Unit 1, which is the oldest nuclear power plant (NPP) in Korea has been operated since 1978. In addition, 10 other NPPs have been operating more than 10 years. As the number of aging plants rise, public concern over the safety of operating NPPs has increased. Periodic safety review (PSR) in addition to the existing safety assessments are proposed by IAEA as an effective way to verify that operating NPPs maintain the high level of safety. In this regard, the Ministry of Science and Technology (MOST), Korea’s nuclear regulatory body, recently established an institutional process through revision to the atomic energy act to introduce PSR. This PSR considers, among other factors, improvements in safety standards and practices, the cumulative effects of plant aging, operating experience, and the evolution of science and technology. In particular, the assessment and management of plant aging is one of the major areas. It includes identification of the system, structure and components (SSCs) for aging management, assessment of aging effects and planning of aging management implementation program. PSR results could be one of the procedural requirements that are utilized to renew an operating license of a NPP. This paper describes safety assessment requirements including PSR and aging management activities in Korea. This paper also includes the strategy and method for the application of PSR results to the aging management and continued operation of NPPs.

Abstract: In order to simulate the growth of arbitrarily shaped three-dimensional cracks, the finite element alternating method is extended. As the required solution for a crack in an infinite body, the symmetric Galerkin boundary element method formulated by Li and Mear is used. In the study, a crack is modeled as distribution of displacement discontinuities, and the governing equation is formulated as singularity-reduced integral equations. With the proposed method several example problems, such as a penny-shaped crack, an elliptical crack in an infinite solid and a semi-elliptical surface crack in an elbow are solved. And their growth under fatigue loading is also considered and the accuracy and efficiency of the method are demonstrated.

Abstract: The multi-pass welding generates residual stress which may change fatigue crack growth rate and impair the lifetime of nuclear welded structures. In this paper, we performed fatigue test with notched specimens and evaluate the effect of residual stress on fatigue crack growth rate of welds. In order to identify the magnitude of residual stress, the residual stress was measured by HDM (hole drilling method) and residual stress analysis was performed by using FEM (finite element method). In order to review the effect of residual stress on fatigue crack growth rate, the fatigue crack growth analysis was also performed to determine the fatigue crack growth curves by using FEM and ASME B&PV Code, Sec.XI, App.A. Finally, as a result of comparison among the fatigue crack growth curves, it is found that the fatigue crack growth rate was quite different according to the crack location even if the residual stresses are considered.

Abstract: Detailed two- and three-dimensional finite element analyses are performed to develop an engineering method to estimate elastic-plastic J for cracked structures under combined primary and secondary stresses, based on the V-factor. Extensive analyses with a wide range of geometry and load combination are considered. The results suggest important factors affecting the value of the V-factor. The most important one is the load magnitude, parameterized by the proximity of plastic yielding. The second one is the relative magnitude of the secondary stress to the primary stress. Although the effect of material, in particular materials with Lüders strain seems to be present, such an effect could be neglected, compared to those of the above two parameters. Based on the present results, an engineering method to estimate J for combined primary and secondary stress can be proposed using bilinear equations in terms of the above two parameters.

Abstract: Reactor pressure vessel (RPV) is the most critical component in nuclear power plant. RPV is subjected to radiation embrittlement, which is characterized as neutron fluence-dependent reduction in fracture toughness of the material. Therefore, risk for potential failure of RPV increases as operating time and fluence level increase. To prevent the potential failure, it is requested for RPV to operate in accordance with pressure-temperature (P-T) limit curve during operation. However, it has been reported that P-T limit curve which is typically developed in accordance with the procedure in ASME code is too conservative. Therefore, in order to investigate the conservatism of current P-T limit curve and develop more realistic one, probabilistic approach based on the risk was utilized in this paper. The resulting P-T limit curve is very higher than that from deterministic approach, and can be used as alternative operation limit of the RPV, because probabilistic P-T limit curve seems to have enough safety margin for potential failure of RPV.

Abstract: Operating experience of steam generators has shown that cracks of various morphologies frequently occur in the steam generator tubes. These cracked tubes can stay in service if it is proved that the tubes have sufficient safety margin to preclude the risk of burst and leak. Therefore, integrity assessment using exact limit load solutions is very important for safe operation of the steam generators. This paper provides global and local limit load solutions for surface cracks in the steam generator tubes. Such solutions are developed based on three-dimensional (3-D) finite element analyses assuming elastic-perfectly plastic material behavior. For the crack location, both axial and circumferential surface cracks, and for each case, both external and internal cracks are considered. The resulting global and local limit load solutions are given in polynomial forms, and thus can be simply used in practical integrity assessment of the steam generator tubes, because the comparison between experimental data and FE solutions shows good agreement.

Abstract: In the paper, the validity of the modified mesh-insensitive SS (structural stress) procedure to apply to the welded joints with local thickness variation is identified via the comparison of SCFs (stress concentration factors) calculated for various FE (finite element) models. FCI (fatigue crac kinitiation) cycles are determined by using the SS/EPFM (elasto-plastic fracture mechanics) approach and the various fatigue crack growth models. Fatigue test is performed to identify the validity of the fatigue analysis results. Finally, as a result of comparison among test and various analysis results, it is found that the SS/FM (fracture mechanics) approach agrees well with the fatigue test results over all cycle regions and the SS/EPFM approach is more reliable than the SS/LEFM (linear elastic fracture mechanics) approach.