• Nuclear Engineering and Technology
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• 제44권8호
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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.

• 한국압력기기공학회 논문집
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• 제15권1호
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.283.1-283
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 추계학술발표회요약집
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• pp.246-246
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• 1998

• Corrosion Science and Technology
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• 제8권5호
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• pp.188-192
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• 2009

A forced outage due to a steam generator tube leak in a Korean nuclear power plant was reported.1) Primary water stress corrosion cracking has occurred in many tubes in the plant, and they were repaired using sleeves or plugs. In order to develop proper repair criteria, it is necessary to understand the leak behavior of the tubes containing stress corrosion cracks. Stress corrosion cracks were developed in 0.1 M sodium tetrathionate solution at room temperature. Steam generator(SG) tubes with short cracks were successfully fabricated with a restricted solution contact method. The leak rates of the degraded tubes were measured at room temperature. Some tubes with 100 % through wall cracks showed an increase of leak rate with time at a constant pressure.

• Nuclear Engineering and Technology
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• 제53권6호
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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.

• Corrosion Science and Technology
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• 제22권1호
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• pp.44-54
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• 2023

The rupture disk corrosion test (RDCT) method was recently developed to evaluate stress corrosion cracking (SCC) and was found to have great potential for the real-time detection of SCC initiation in a high temperature and pressure environment, simulating the primary water coolant of pressurized water reactors. However, it is difficult to directly measure the stress applied to a disk specimen, which is an essential factor in SCC initiation. In this work, finite element analysis (FEA) was performed using ABAQUS^TM to calculate the stress and deformation of a disk specimen. To determine the best mesh design for a thin disk specimen, hexahedron, hex-dominated, and tetrahedron models were used in FEA. All models revealed similar dome-shaped deformation behavior of the disk specimen. However, there was a considerable difference in stress distribution in the disk specimens. In the hex-dominated model, the applied stress was calculated to be the maximum at the dome center, whereas the stress was calculated to be the maximum at the dome edge in the hexahedron and tetrahedron models. From a comparison of the FEA results with deformation behavior and SCC location on the disk specimen after RDCT, the most proper FE model was found to be the tetrahedron model.

• 대한기계학회논문집A
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• 제28권6호
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• pp.732-738
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• 2004

One of the main degradation mechanisms in steam generator tubes is stress corrosion cracking induced by residual stress. The resulting damages can cause tube bursting or leakage of the primary water which contains radioactivity. Shot peening technique has been used to prevent stress corrosion crack growth in steam generator tubes. In order to investigate the shot peening effect on stress corrosion cracking stress intensity factors are calculated for the semi-elliptical surface crack which is located in residual stress region. The residual stress distribution in steam generator tubes is obtained from the simple model proposed by Frederick et al.

• Nuclear Engineering and Technology
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• 제49권4호
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• pp.817-824
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• 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.

• 한국압력기기공학회 논문집
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• 제7권1호
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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.