• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.20-27
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• 2010

In order to prevent the corrosion of component contacted primary water designed alloy 600 material in the nuclear power plant. But the primary water stress corrosion cracking(PWSCC) of alloy 600 and weld area occurs continuously due to the residual stress. The leakage accident resulted from PWSCC in the drain nozzle of the steam generator of domestic power plants. Heater sleeves of the pressurizer are welded with alloy 600 weld material and therefore exposed to the primary water environment. PWSCC occurred in heater sleeve material and weld area of many foreign power plants. The current issue of domestic nuclear power plants are consequently concentrated to PWSCC of similar material. In order to improve the detection and the sizing of the PWSCC in the welding sleeve of the pressurizer, the automatic UT system and multi-directions probe sets have been developed. The experimental studies have been performed using the mock-up block containing artificial reflectors(ID connected EDM notch) and semi-artificial cracks made from thermal fatigue. The automatic UT System is applied in the detection and the length sizing of the ID/OD on the tube and the J-groove weld area of the artificial reflectors and results of the detection and the sizing are compared respectively. Also, the developed automatic UT system is successfully accomplished to inspect the heater sleeve and the J-groove weld area on the pressurizer for the detection of PWSCC.

• Nuclear Engineering and Technology
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• v.30 no.5
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• pp.435-443
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• 1998

Leak-before-break(LBB) approach has been shown to be both cost effective and risk reductive when applied to high energy Piping in nuclear Power Plants. For the Korean Next Generation Reactor (KNGR) development, LBB application is considered for the Main Steam Line(MSL) piping inside containment. Unlike the primary system leakages, the MSL leak detection systems must be based on principles other than radioactivity measurements. Among humidity, heat and acoustic noise currently being considered as indicators of leakage, we explored humidity as an effective one and developed ceramic-based humidity sensor which can be qualified for LBB applications. The ceramic material, sintered and annealed MgCr$_2$O$_4$-TiO$_2$, is shown to increase its electrical conductivity drastically upon water vapor adsorption over the entire temperature range of interest. With this ceramic sensor specimen, we suggested installation-inside-the-piping method by which we can detect leakage more rapidly and sensitively. In this paper, we describe the progress in the development and characterization of ceramic humidity sensor for the LBB application to the MSL of KNGR.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.1
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• pp.75-83
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• 2017

This study investigated deformation behavior of major nuclear structural materials under cyclic loading conditions via cyclic stress-strain test. The cyclic stress-strain tests were conducted on SA312 TP316 stainless steel and SA508 Gr.3 Cl.1 low-alloy steel, which are used as materials for primary piping and reactor pressure vessel nozzle respectively, under cyclic load with constant strain amplitude and constant load amplitude at room temperature (RT) and $316^{\circ}C$. From the results of tests, the cyclic hardening and softening behavior, stabilized cyclic stress-strain behavior, and ratcheting behavior of both materials were investigated at both RT and $316^{\circ}C$. In addition, appropriate considerations for cyclic deformation behavior in the structural integrity evaluation of major nuclear components under excessive seismic condition were discussed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.1
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• pp.40-45
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• 2019

Unlike the welded pipes in the primary system of light water nuclear power plants being periodically inspected with in-Service inspection program, relatively small pipes with the outer diameter less than 2 inch have not been regularly inspected to date. However, after several failure reports on the occurrence of critical crack-like defects in small pipes, inspection for the small pipes has been more demanded because it could cause the provisional outage of nuclear power plants. Nevertheless, there's no particular method to examine the small pipes having access limitations for inspection due to various reasons; inaccessible area, excessive radiation exposure, hazardous surrounding, and etc. This study is to develop a reliable inspection technique using torsional and flexural modes of guided wave to detect defects that could occur in inaccessible area. The attribute of guided wave that can travel a long distance enables to inspect even isolated range of the pipe from accessible location. This paper presents a case study of the evaluation test on 3/4" small-bore pipes with guide wave method. The test result demonstrates the crack signal behavior and assures possibility to detect the crack signal in a flexural mode, which is clearly distinguishable from the symmetric structure signal in a torsional mode.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.30-40
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• 2016

The paper investigates the previous studies about welding residual stresses in nuclear components. First, various residual stress measurement methods are reviewed in applicability. Second a finite element welding residual stress analysis technique, which was developed from the viewpoint of FFS (Fitness-For-Service) assessment, is explained. Third, characteristics of the welding residual stresses on J-groove welds and butt welds were presented via investigating the previous studies. Last, engineering formulae for residual stresses in the FFS assessment codes such as R6 and API 579/ASME FFS-1 Code is summarized.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.17-22
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• 2016

Alloy 600 widely used in nuclear power plant is susceptible to primary water stress corrosion cracking (PWSCC). It is important to prioritize the inspection of Alloy 600 components using PWSCC susceptibility index. Plant-specific model for the susceptibility index was reviewed. The normalized PWSCC susceptibility index to a reference value is suggested and applied. The result was found to be reasonable.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.1
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• pp.25-31
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• 2009

Steam Generator(SG) tube is an important component of Nuclear Power Plant(NPP), which comprises of the pressure boundary of primary system. The integrity of SG tube has been confirmed by the eddy current test every outage. In Korea, Bobbin probe and MRPC probe have been generally used for the eddy current test. Meanwhile the usage of Array probe has gradually increased in U.S., Japan and other countries. In this study, we investigated the defect detection capability of the Array probe through its preliminary application to SG tube inspection. The Array probe has the equivalent capability in the defect detection and sizing as the conventional methods. Thus it is desirable that the Array probe is generally applied to SG tube inspection in the domestic NPPs.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.1
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• pp.19-24
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• 2012

This paper presents the results of finite element analysis for the effects of hydrostatic test and operating condition on the weld residual stress at dissimilar metal weld of a pressurizer safety nozzle in a nuclear power plant. For the study, the weld residual stress at ambient condition was analyzed using ABAQUS in the first place. After the weld residual stress analysis, the hydrostatic test condition and operating condition was applied to the same model one after another. The weld residual stress was observed to change due to the successive hydrostatic test and operating condition. The axial residual stresses on inner surface of the dissimilar metal weld and HAZ region were decreased by hydrostatic test and operating condition, which gives beneficial effect on preventing primary water stress corrosion cracking.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.131-136
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• 2003

CF8M cast austenitic stainless steel is used for several components such as primary coolant piping, elbow, pump casing, and valve bodies in light water reactors. These components are subject to thermal aging at the reactor operating temperature. Thermal aging results in spinodal decomposition of the delta-ferrite leading to increased strength and decreased toughness. In this study, three kinds of the aged CF8M specimen were prepared using an artificially simulated aging method. The objective of this study is to summarize the method of estimating ferrite contents, Charpy impact energy and J-R curve, and to evaluate the thermal embrittlement of the CF8M cast austenitic stainless steel piping used in the domestic nuclear power plants.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.2
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• pp.19-30
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• 2017

Austenitic Stainless Steels (ASSs) are widely used as structural materials in the pressurized water reactors (PWRs) because of their superior mechanical properties and corrosion resistance. However, it is well known that ASSs are susceptible to the environmental assisted cracking (EAC) such as environmental assisted fatigue (EAF) during the long term operation. There have been extensive tests and researches to understand the extent and the mechanisms of environmental effects. In this paper, the world-wide EAF test results of ASSs are introduced including those of Korean test programs. The suggested EAF mechanisms of ASSs are also discussed. Finally, the areas of further research to resolve the issue of EAF are suggested.