• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.343-354
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• 2012

Primary water stress corrosion cracking (PWSCC) is a major safety concern in the nuclear power industry worldwide. PWSCC is known to initiate only in the condition in which sufficiently high tensile stress is applied to alloy 600 tube material or alloy 82/182 weld material in pressurized water reactor operating environments. However, it is still uncertain how much tensile stress is re-quired to generate PWSCC or what causes such high tensile stress. This study was performed to pre-dict the magnitude of weld residual stress and operating stress and compare it with previous experi-mental results for PWSCC initiation. For the study, a pressurizer safety nozzle was selected because it is reported to be vulnerable to PWSCC in overseas plants. The assessment was conducted by nu-merical analysis. Before performing stress analysis for plant conditions, a preliminary mock-up ana-lysis was done. The result of the preliminary analysis was validated by residual stress measurement in the mock-up. After verification of the analysis methodology, an analysis under plant conditions was conducted. The analysis results show that the stress level is not high enough to initiate PWSCC. If a plant is properly welded and operated, PWSCC is not likely to occur in the pressurizer safety nozzle.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.1036-1046
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• 2016

We propose a primary water stress corrosion cracking (PWSCC) initiation model of Alloy 600 that considers the stress triaxiality factor to apply to finite element analysis. We investigated the correlation between stress triaxiality effects and PWSCC growth behavior in cold-worked Alloy 600 stream generator tubes, and identified an additional stress triaxiality factor that can be added to Garud's PWSCC initiation model. By applying the proposed PWSCC initiation model considering the stress triaxiality factor, PWSCC growth simulations based on the macroscopic phenomenological damage mechanics approach were carried out on the PWSCC growth tests of various cold-worked Alloy 600 steam generator tubes and compact tension specimens. As a result, PWSCC growth behavior results from the finite element prediction are in good agreement with the experimental results.

• Corrosion Science and Technology
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• v.18 no.1
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• pp.33-38
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• 2019

Alloy 600/182 with excellent mechanical/chemical properties have been utilized for nuclear power plants. Although both alloys are known to have superior corrosion resistance, stress corrosion cracking failure has been an issue in primary water environment of nuclear power plants. Therefore, primary water stress corrosion crack (PWSCC) growth rate tests were conducted to investigate crack growth properties of Alloy 600/182. To investigate PWSCC growth rate, test facilities including water chemistry loop, autoclave, and loading system were constructed. In PWSCC crack growth rate tests, half compact-tension specimens were manufactured. These specimens were then placed inside of the autoclave connected to the loop to provide primary water environment. Tested conditions were set at temperature of $360^{\circ}C$ and pressure of 20MPa. Real time crack growth rates of specimens inside the autoclave were measured by Direct Current potential drop (DCPD) method. To confirm inter-granular (IG) crack as a characteristic of PWSCC, fracture surfaces of tested specimens were observed by SEM. Finally, crack growth rate was derived in a specific stress intensity factor (K) range and similarity with overseas database was identified.

• Nuclear Engineering and Technology
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• v.44 no.8
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• pp.961-970
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• 2012

Often, probabilistic fracture mechanics (PFM) approaches have been adopted to quantify the failure probabilities of Ni-base alloy components, especially due to primary water stress corrosion cracking (PWSCC), in a primary piping system of pressurized water reactors. In this paper, the key features of an advanced PFM code, PINEP-PWSCC (Probabilistic INtegrity Evaluation for nuclear Piping-PWSCC) for such purpose, are described. In developing the code, we adopted most recent research results and advanced models in calculation modules such as PWSCC crack initiation and growth models, a performance-based probability of detection (POD) model for Ni-base alloy welds, and so on. To verify the code, the failure probabilities for various Alloy 182 welds locations were evaluated and compared with field experience and other PFM codes. Finally, the effects of pre-existing crack, weld repair, and POD models on failure probability were evaluated to demonstrate the applicability of PINEP-PWSCC.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1909-1923
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• 2021

We experimentally investigated the effects of thermal aging and cold work on the microstructure, mechanical properties, and primary water stress corrosion cracking (PWSCC) initiation time for Alloy 182 welds. The effects of thermal aging and cold work on the PWSCC initiation time of Alloy 182 were modeled based on the plastic energy concept and the PWSCC initiation data of this study and previous reports by considering censored data. Based on the results, it is estimated that the PWSCC resistance of the Alloy 182 weld firstly increases and then decreases with thermal aging time when the applied stress is kept constant.

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

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.1
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• pp.28-32
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• 2019

In nuclear power plant operation and spent fuel canisters, it is necessary to provide a sound technical basis for the safety and security of long-term operation and storage respectively. Recently, the peening technology is being discussed and the technology will be adopted to ASME Section III, Division 1, Subsection NX (2019 Edition). The peening is prohibited in current edition, but it will be approved in 2019 Edition and adopted. However, Surface stress improvement techniques such as the peening is used to mitigate SCC susceptible in operating nuclear plants. Although the peening will be approved to ASME CODE, there are no performance criteria listed in the 2019 edition. The Korean International Working Group (KIWG) formed a new Task Group named "Advanced Surface Stress Improved Technology". The task group will develop a CODE CASE to address PWSCC(Primary Water Stress Corrosion Cracking) and CISCC(Chloride Induced Stress Corrosion Cracking) for new ASME Section III components. TG-ASSIT was started to make peening performance criteria for ASME Section III (new fabrication) applications. The objective of TG-ASSIT is to gain consensus among the relevant Code groups that requirements/mitigation have been met.

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

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.25 no.10
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• pp.568-572
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• 2013

The main components, such as a reactor vessel, in commercial nuclear power plants have been welded to pipes with dissimilar metal in which Primary Water Corrosion Cracking (PWSCC) has been occurred. PWSCC has become a worldwide issue recently. This paper addresses the results of experimental and numerical analysis to prevent PWSCC by changing the stress profile that is tensile stress to compressive stress at interesting regions with plastic deformation generated by mechanical pressure. Based on the results of experimental and numerical analysis with a 6 inch pipe and dissimilar metal welded pipes, compressive stress 68~206 Mpa is generated at all locations of inner surface in the heat affected zone.

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