• Proceedings of the Korea Environmental Mutagen Society Conference
• /
• 2004.11a
• /
• pp.104-104
• /
• 2004

• Corrosion Science and Technology
• /
• v.22 no.4
• /
• pp.265-272
• /
• 2023

The aim of this study was to develop a new in-situ observation method and instrument in micro-scale to investigate the mechanism of stress corrosion cracking (SCC) initiation of Ni-base alloys in a high temperature water environment of pressurized water reactors (PWRs). A laser confocal microscope (LCM), an autoclave with diamond window view port, and a slow strain-rate tester with primary water circulation loop system were components of the instrument. Diamond window, one of the core components of the instrument, was selected based on its optical, chemical, and mechanical properties. LCM was used to observe the specimen in micro-scale, considering the experimental condition of a high-temperature primary water environment. Using in-situ method and instrument, it is possible to observe oxidation and deformation of specimen surface in micro-scale through the diamond window in a high-temperature primary water in real-time. The in-situ method and instrument developed in this work can be utilized to investigate effects of various factors on SCC initiation in a high-temperature water environment.

• Proceedings of the KSME Conference
• /
• 2004.11a
• /
• pp.7-12
• /
• 2004

A lifetime prediction of holddown spring screw in nuclear fuel assembly was performed using fracture mechanics approach. The spring screw was designed such that it was capable of sustaining the loads imposed by the initial tensile preload and operational loads. In order to investigate the cause of failure and to predict the stress corrosion cracking life of the screw, a stress analysis of the top nozzle spring assembly was done using finite element analysis. The elastic-plastic finite element analysis showed that the local stresses at the critical regions of head-shank fillet and thread root significantly exceeded than the yield strength of the screw material, resulting in local plastic deformation. Normalized stress intensity factors for PWSCC life prediction was proposed. Primary water stress corrosion cracking life of the Inconel 600 screw was predicted by using integration of the Scott model and resulted in 1.78 years, which was fairly close to the actual service life of the holddown spring screw.

• Journal of the Korean Society of Safety
• /
• v.9 no.4
• /
• pp.29-41
• /
• 1994

High temperature tensiles tests, steady state creep tests, internal stress tests and creep rupture tests using Al 7075 alloy were performed over the temperature range of 9$0^{\circ}C$~50$0^{\circ}C$ and stress range of 0.64~17.2(kgf/$\textrm{mm}^2$) in order to investigate the creep behavior and predict creep rupture life From the apparent activation energy Qc and the applied stress exponent n measured, at the temperature range of 9$0^{\circ}C$~l2$0^{\circ}C$, the creep deformation seemed to be controlled by cross slip. On the other hand at the temperature of 20$0^{\circ}C$~23$0^{\circ}C$ the creep deformation seemed to be controlled by dislocation climb but at 47$0^{\circ}C$~50$0^{\circ}C$, by diffusion creep. And the rupture life(t$_{f}$) might be represented by anthermal process attributed to the difference of the applied stress dependence of Internal stress and the ratio of the Internal stress to the applied stress, the thermal activated process attributied to the temperature dependence of the internal stress. Also the ratio between stress dependence of primary creep rate and that of minimum creep rate was measured 0.46, the minimum creep rate is expected to be appromately obtained from master creep curve including the relationship primary creep rate and minumum creep rate. Finally the relationship new rupture parameter and logarithmic stress was represented with including the ratio between the dependence of primary creep rate and that of minimum creep rate, using the new rupture parameter the rupture life predition is exactly expected.d.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.32 no.8
• /
• pp.660-667
• /
• 2008

During the last two decades, lots of efforts have been devoted to resolve thermal stratification phenomenon and primary water environment issues. While several effective methods were proposed especially in related to thermally stratified flow analyses and corrosive material resistance experiments, however, lack of details on specific stress and fatigue evaluation make it difficult to quantify structural behaviors. In the present work, effects of the thermal stratification and primary water are numerically examined from a structural integrity point of view. First, a representative austenitic nuclear piping is selected and its stress components at critical locations are calculated in use of four stratified temperature inputs and eight transient conditions. Subsequently, both metal and environmental fatigue usage factors of the piping are determined by manipulating the stress components in accordance with NUREG/CR-5704 as well as ASME B&PV Codes. Key findings from the fatigue evaluation with applicability of pipe and three-dimensional solid finite elements are fully discussed and a recommendation for realistic evaluation is suggested.

• Nuclear Engineering and Technology
• /
• v.47 no.7
• /
• pp.895-906
• /
• 2015

Background: Mitigation of primary water stress corrosion cracking (PWSCC) is a significant issue in the nuclear industry. Advanced nickel-based alloys with lower susceptibility have been adopted, although they do not seem to be entirely immune from PWSCC during normal operation. With regard to structural integrity assessments of the relevant components, an accurate evaluation of crack growth rate (CGR) is important. Methods: For the present study, the extended finite element method was adopted from among diverse meshless methods because of its advantages in arbitrary crack analysis. A user-subroutine based on the strain rate damage model was developed and incorporated into the crack growth evaluation. Results: The proposed method was verified by using the well-known Alloy 600 material with a reference CGR curve. The analyzed CGR curve of the alternative Alloy 690 material was then newly estimated by applying the proven method over a practical range of stress intensity factors. Conclusion: Reliable CGR curves were obtained without complex environmental facilities or a high degree of experimental effort. The proposed method may be used to assess the PWSCC resistance of nuclear components subjected to high residual stresses such as those resulting from dissimilar metal welding parts.

• The Transactions of the Korean Institute of Power Electronics
• /
• v.24 no.6
• /
• pp.406-410
• /
• 2019

Three-level (3L) DC-DC converters are appropriate for high-input-voltage applications. Although the voltage stress of TL converter switches can be reduced to half of the input voltage, the primary side has a large circulating current, which degrades efficiency. In this study, a dual half-bridge cascaded TL converter is presented to reduce this circulating current and thus decrease the conduction loss of the primary circuit. Moreover, the proposed converter can reduce the voltage stress of rectifier diodes, thereby reducing their conduction loss. Therefore, efficiency can be improved by reducing the conduction loss of the primary circuit and rectifier diodes.

• Journal of the Korean Society for Precision Engineering
• /
• v.16 no.3 s.96
• /
• pp.145-153
• /
• 1999

This study has been performed to investigate the stress distribution around defects that behave as stress concentrators and fracture mechanical analysis for cracks initiatiating at stress concentrators. The stress distribution was analyzed using Finite Element Method and non dimensional stress intensity factor was determined by the mean stress method. In addition, stress interaction effects around defects and cracks were compared.

• Nuclear Engineering and Technology
• /
• v.49 no.4
• /
• pp.817-824
• /
• 2017

Residual stress is a critical element in determining the integrity of parts and the lifetime of welded structures. It is necessary to estimate the residual stress of a welding zone because residual stress is a major reason for the generation of primary water stress corrosion cracking in nuclear power plants. That is, it is necessary to estimate the distribution of the residual stress in welding of dissimilar metals under manifold welding conditions. In this study, a cascaded support vector regression (CSVR) model was presented to estimate the residual stress of a welding zone. The CSVR model was serially and consecutively structured in terms of SVR modules. Using numerical data obtained from finite element analysis by a subtractive clustering method, learning data that explained the characteristic behavior of the residual stress of a welding zone were selected to optimize the proposed model. The results suggest that the CSVR model yielded a better estimation performance when compared with a classic SVR model.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.7 no.1
• /
• pp.9-16
• /
• 2011

Primary water stress corrosion cracking (PWSCC) has been found in the weld region of the nuclear power plant. Welding can produce tensile residual stress. Tensile residual stress contributes to the initiation and growth of PWSCC. It is important to estimate weld residual stress accurately to predict or prevent the initiation and growth of PWSCC. This paper shows the results of finite element analysis and measurement experiment for weld residual stress. For the study, four kinds of specimen were fabricated with the materials used in the nuclear power plant. Residual stresses were measured by four kinds of methods of hole drilling, x-ray diffraction, instrumented indentation and sectioning. Through the study, numerical analysis and measurement results were compared and the characteristics of each measurement technique were observed.