• Corrosion Science and Technology
• /
• v.14 no.3
• /
• pp.113-119
• /
• 2015

Alloy 82/182 weld metals had been extensively used in joining the components of the PWR primary system. Unfortunately, there have been a number of incidents of cracking caused by PWSCC in Alloy 82/182 welds during the operation of PWR worldwide. To mitigate PWSCC, optimization of water-chemistry conditions, especially dissolved hydrogen (DH) and Zn contents, is considered as the most promising and effective remedial method. In this study, the PWSCC behaviours of Alloy 182 weld were investigated in simulated PWR environments with various DH content. Both in-situ and ex-situ oxide characterizations as well as PWSCC initiation tests were performed. The results showed that PWSCC crack initiation time was shortest in PWR water (DH: 30cc/kg). Also, high stress reduced crack initiation time. Oxide layer showed multi-layered structures consisted of the outer needle-like Ni-rich oxide layer, Fe-rich crystalline oxide, and inner Cr-rich inner oxide layers, which was not altered by the level of applied stress. To analyse the multi-layer structure of oxides, EIS measurement were fitted into an equivalent circuit model. Further analyses including TEM and EDS are underway to verify appropriateness of the equivalent circuit model.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.37 no.1
• /
• pp.47-54
• /
• 2013

This study investigates the crack growth behavior due to primary water stress corrosion cracking (PWSCC) in the dissimilar metal butt weld of a reactor piping using Alloy 82/182. First, detailed finite element stress analyses were performed to predict the stress distribution of the dissimilar metal butt weld in which the hydrostatic and the normal operating loads as well as the weld residual stresses were considered to evaluate the stress redistribution due to mechanical loadings. Based on the stress distributions along the wall thickness of the dissimilar metal butt weld, the crack growth behavior of the postulated axial and circumferential cracks were predicted, from which the crack growth diagram due to PWSCC was proposed. The present results can be applied to predict the crack growth rate in the dissimilar metal butt weld of reactor piping due to PWSCC.

• Nuclear Engineering and Technology
• /
• v.51 no.7
• /
• pp.1805-1815
• /
• 2019

While nickel-based alloys have been widely used for power plants due to corrosion resistance and good mechanical properties, during the last couple of decades, failures of nuclear components increased gradually. One of main degradation mechanisms was primary water stress corrosion cracking at dissimilar metal welds of piping and reactor head penetrations. In this context, precise estimation of welding effects became an important issue for ensuring reliability of them. The present study deals with a series of finite element analyses and crack growth rate evaluation of Alloys 690/152. Firstly, variation of residual stresses and equivalent plastic strains was simulated taking into account welding of a cylindrical block. Subsequently, extraction and pre-cracking of compact tension (CT) specimens were considered from different locations of the block. Finally, crack growth curves of the alloys and heat affected zone were developed based on analyses results combined with experimental data in references. Characteristics of crack growth behaviors were also discussed in relation to mechanical and fracture parameters.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.32 no.10
• /
• pp.873-881
• /
• 2008

Weld overlay is one of the residual stress mitigation methods which arrest crack initiation and crack growth. Therefore weld overlay can be applied to the region where cracking is likely to be. An overlay weld used in this manner is termed a preemptive weld overlay(PWOL). In pressurized water reactor(PWR) dissimilar metal weld is susceptible region for primary water stress corrosion cracking(PWSCC). In order to examine the effect of PWOL on residual stress mitigation, PWOL was applied to a specific dissimilar metal weld of Kori nuclear power plant by finite element analysis method. As a result, strong compressive residual stress was made in PWSCC susceptible region and PWOL was proved effective preemptive repair method for weldment.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.32 no.9
• /
• pp.770-781
• /
• 2008

In nuclear power plants, ferritic low alloy steel components were connected with austenitic stainless steel piping system through alloy 82/182 butt weld. There have been incidents recently where cracking has been observed in the dissimilar metal weld. Alloy 82/182 is susceptible to primary water stress corrosion cracking. Weld-induced residual stress is main factor for crack growth. Therefore exact estimation of residual stress is important for reliable operating. This paper presents residual stress computation performed by 6" safety & relief nozzle. Based on 2 dimensional and 3 dimensional finite element analyses, effect of welding variables on residual stress variation is estimated for sensitivity analysis.

• Journal of Welding and Joining
• /
• v.27 no.4
• /
• pp.32-37
• /
• 2009

Alloy 600 and Alloy 82/182 materials have been used widely in PWR plants. But these materials are known to be susceptible to PWSCC(Primary Water Stress Corrosion Cracking). Recently, there have been several PWSCC events in major components due to repair welding, because repair welding in the dissimilar metal welds during the construction increases residual stress significantly on the inner surface of welds. In this paper, various residual stress analyses for repair welding were performed using FEM to check the effect of repair welding on residual stress distributions in PZR safety/relief nozzle. The results indicate that for inside surface repair welding, high tensile residual stress is developed on the inside surface of the nozzles.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.21 no.5
• /
• pp.501-509
• /
• 2001

For the enhancement of ECT reliability on the primary water stress corrosion cracks of nuclear steam generator tubes, of which the occurrence is on the increase, it is important to comprehend the signal characteristics on crack morphology and to select an appropriate probe type. In this paper, the sizing accuracy and the detectability for the inner wall axial cracks of tubes were quantitatively evaluated using the following specimens: the electric discharge machined notches and the corrosion cracks which were developed on the operating steam generator tubes. The difference of eddy current signal characteristics between pancake and axial coil were also Investigated. The results obtained from this study provide a useful information for more precise evaluation on the inner wall axial tracks oi stram generator tubes.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.39 no.3
• /
• pp.269-273
• /
• 2015

The study investigated the effect of the distance between two collinear circumferential surface cracks on the primary stress corrosion crack (PWSCC) growth in alloy 600TT steam generator tubes using a finite element damage analysis based on the PWSCC initiation model and macroscopic phenomenological damage mechanics approach. The damage analysis method was verified by comparing the results to the previous study results. The verified method was applied to collinear circumferential surface PWSCCs. As a result, it was found that the collinear cracks showed earlier coalescence and penetration times than the a single crack, and the times increased with the distance. In addition, it is expected that penetration may occur before coalescence of two cracks if they are more than a specific distance apart.

• Nuclear Engineering and Technology
• /
• v.49 no.7
• /
• pp.1524-1536
• /
• 2017

During the fabrication process of reactor vessel head penetration (RVHP), the grain size of the tube material can be changed by hot or cold work and the inner side of the tube can also be shrunk due to welding outside of the tube. Several nonregular time-of-flight diffraction (TOFD) signals were found because of deformed grains. In this paper, an investigation of nonregular TOFD indications acquired from RVHP tubes using experiments and computer simulation was performed in order to identify and distinguish TOFD signals by coarse grains from those by Primary Water Stress Corrosion Crack (PWSCC). For proper understanding of the nonregular TOFD indications, microstructural analysis of the RVHP tubes and prediction of signals scattered from the grains using Finite Element Method (FEM) simulation were performed. Prediction of ultrasonic signals from the various sizes of side drilled holes to find equivalent flaws, determination of the size of the nonregular TOFD indications from the coarse grains, and experimental investigation of TOFD signals from coarse grain and shrinkage geometry to identify PWSCC signals were performed. From the computer simulation and experimental investigation results, it was possible to obtain the nonregular TOFD indications from the coarse grains in the alloy 690 penetration tube of RVHP; these nonregular indications may be classified as PWSCC. By comparing the computer simulation and experimental results, we were able to confirm a clear difference between the coarse grain signal and the PWSCC signal.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.36 no.9
• /
• pp.1047-1052
• /
• 2012

Primary water stress corrosion cracking (PWSCC) instances have been reported in the Alloy 600 reactor pressure vessel head penetration nozzle and the Alloy 82/182 dissimilar metal butt weld nozzle in several PWRs. Therefore, in-service inspection programs have been adopted worldwide to prevent failure at the weld region. If a PWSCC is observed at the dissimilar metal weld region during inspection, its structural integrity should be evaluated; however, this requires considerable time and effort, and this might lead to a decrease in the plant utilization coefficient. To prevent this, KHNP-CRI have established integrity assessment criteria and developed a computer program for the fast evaluation and judgment of PWSCC. In this paper, the results and current status of the same are presented. Through this study, criteria for the structural integrity evaluation of PWSCC have been established, and a computer program has been developed to realize technical means for the evaluation of PWSCC structural integrity.