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Yamaguchi, Yoshihito; Mano, Akihiro; Li, Y.
Transactions of 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 10 Pages, 2024/03
The steam generator (SG) is an important component of a pressurized water reactor. In addition, local wall-thinning has been reported in SG tubes. The burst differential pressure, considering both the internal and external pressures from the primary and secondary coolant systems, should be predicted for the failure probability evaluation or structural integrity assessment of SG tubes. In this study, based on the results of burst tests performed in Japan and the United States, we improved the existing burst pressure estimation method for SG tubes with wall-thinning. In addition, as an example of the utilization of the improved burst pressure estimation method, the conditional failure probabilities for SG tubes with local wall-thinning, which is necessary for probabilistic risk assessment and risk-informed decision making, are calculated considering the dimensions of the wall-thinning.
Li, S.; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.; Deng, D.*
Proceedings of ASME 2023 Pressure Vessels and Piping Conference (PVP 2023) (Internet), 7 Pages, 2023/07
Yamaguchi, Yoshihito; Takamizawa, Hisashi; Katsuyama, Jinya; Li, Y.
Proceedings of ASME 2023 Pressure Vessels and Piping Conference (PVP 2023) (Internet), 9 Pages, 2023/07
The stress intensity factor (SIF) for crack at nozzle corner is a key parameter in structural integrity assessment of nozzle in reactor pressure vessel (RPV). Although various SIF solutions for surface cracks at nozzle corners have been proposed, most of them are only focusing on the deepest point of the crack, and the information about geometric dimension of the nozzle corner is not clear. According to the previous fatigue test results regarding the surface crack at the nozzle corner, the amounts of crack growth at the surface points were larger than that at the deepest point of the crack. Such results imply that SIFs at the surface points may be higher than that at the deepest point. To increase the reliability of the structural integrity assessment, it is necessary to provide SIF solutions for both surface and deepest points. In this study, SIF solutions for two surface points and the deepest point of surface crack at nozzle corners are developed through finite element analyses and the solutions are provided corresponding to the geometric dimensions of nozzle corner and crack size.
Lacroix, V.*; Hasegawa, Kunio; Li, Y.; Yamaguchi, Yoshihito
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 7 Pages, 2022/07
Yamaguchi, Yoshihito; Hasegawa, Kunio; Li, Y.; Lacroix, V.*
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 4 Pages, 2022/07
Yamaguchi, Yoshihito; Nishida, Akemi; Li, Y.
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 7 Pages, 2022/07
The wall-thinning is one of the most important age-related degradation phenomena in nuclear piping systems. Furthermore, in recent years, several nuclear power plants in Japan have experienced severe earthquakes. Therefore, failure probability analysis and fragility evaluation of piping systems, taking both wall-thinning and seismic response stresses into consideration, have become increasingly important in seismic probabilistic risk assessment. In Japan Atomic Energy Agency, in order to evaluate the failure probability of aged piping system with wall-thinning, a probabilistic analysis code PASCAL-EC was developed. In this study, to evaluate the seismic fragility of a wall-thinned pipe, a model of seismic response stress considering the wall-thinning effect, a failure evaluation method for wall-thinned pipes, and functions related to uncertainties treatment for important influence parameters have been introduced to PASCAL-EC. In this paper, the improved PASCAL-EC is outlined and preliminary results of the seismic fragility evaluation performed using this code are provided.
Yamaguchi, Yoshihito; Mano, Akihiro; Li, Y.
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 10 Pages, 2022/07
The steam generator (SG) tube is one of the important components in pressurized water reactors. Flaws such as wall-thinning or stress corrosion cracking have been reported in SG tubes. The burst pressure where both the internal and external pressures from the primary and secondary coolant systems are considered must be predicted to assess the structural integrity of SG tubes. Burst tests were performed by various organizations. On the basis of the test results, failure estimation methods were proposed. In this study, previous burst test data and existing failure estimation methods for SG tubes with wall-thinning or crack were investigated. As a result, the coefficient of the existing estimation method for SG tube with uniform wall-thinning was updated. In addition, failure estimation methods that are suitable for SG tubes with crack or local wall-thinning were proposed by considering the effects of the flaw shape and size on the burst pressure. The applicability of the failure estimation methods was confirmed by comparing the predicted results with the burst test data in actual SG tubes.
Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Furuta, Takuya; Kaji, Yoshiyuki
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 9 Pages, 2022/07
Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.
Journal of Pressure Vessel Technology, 144(1), p.011506_1 - 011506_9, 2022/02
Times Cited Count：1 Percentile：20.34(Engineering, Mechanical)In the past few decades, the cracks because of stress corrosion cracking (SCC) have been detected in the dissimilar weld joints welded using nickel based alloy in piping system of boiling water reactors. Thus, the structural integrity assessment for such weld joints has become important. Nowadays, probabilistic fracture mechanics (PFM) analysis is recognized as a rational method for structural integrity assessment because it can consider inherent uncertainties of various influencing factors as probability distributions and quantitatively evaluate the failure probability of a cracked component. The Japan Atomic Energy Agency has developed a PFM analysis code PASCAL-SP for a probabilistic structural integrity assessment of weld joint in pipe in nuclear power plant. This study improves the analysis functions of PASCAL-SP for weld joint welded using nickel based alloy in boiling water reactor susceptible to SCC. As an analysis example of the improved version of PASCAL-SP, the failure probability of a weld joint is quantitatively evaluated. Furthermore, sensitivity analyses are conducted concerning the effect of leak detection and in-service inspection. From the analysis results, it is concluded that the improved version of PASCAL-SP is useful for structural integrity assessment.
Zhang, T.; Lu, K.; Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.
Fatigue & Fracture of Engineering Materials & Structures, 44(12), p.3399 - 3415, 2021/12
Times Cited Count：14 Percentile：81.94(Engineering, Mechanical)The Gurson-Tvergaard-Needleman (GTN) model is considered a promising approach in failure prediction as it takes the micromechanical behavior of ductile metals into consideration and its function exhibits a relatively clear physical meaning. Although the GTN model has been widely investigated in the past decades, its engineering applications have scarcely progressed due to the difficulty in determining the eight strongly coupled parameters. Based on the physical background of GTN model, a set of methods was established to determine the parameters in the GTN model. The knowledge of continuum damage mechanics was used to experimentally determine the development of void volume fraction through the variation of effective Young's modulus in a uniaxial tensile test, and three parameters regarding void nucleation were analytically derived using a newly established method. Other parameters in the GTN model were also uniquely determined through a joint use of the chemical composition analysis (for the initial void volume fraction), the cell model analyses (for the two constitutive parameters), and the inverse finite element method (for the two failure parameters). The reliability of this novel parameter determination method was verified through the failure prediction of both cracked and uncracked specimens of carbon steel STPT410.
Yamaguchi, Yoshihito; Li, Y.
Haikan Gijutsu, 63(12), p.22 - 27, 2021/10
no abstracts in English
Yamaguchi, Yoshihito; Katsuyama, Jinya; Masaki, Koichi*; Li, Y.
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 9 Pages, 2021/07
The seismic probabilistic risk assessment is an important methodology to evaluate the seismic safety of nuclear power plants. In this assessment, the core damage frequency is evaluated from the seismic hazard, seismic fragilities, and accident sequence. Regarding the seismic fragility evaluation, the probabilistic fracture mechanics can be applied as a useful evaluation technique for aged piping systems with crack or wall thinning due to the age-related degradation mechanisms. In this study, to advance seismic probabilistic risk assessment methodology of nuclear power plants that have been in operation for a long time, a guideline on the seismic fragility evaluation of the typical aged piping systems of nuclear power plants has been developed considering the age-related degradation mechanisms. This paper provides an outline of the guideline and several examples of seismic fragility evaluation based on the guideline and utilizing the probabilistic fracture mechanics analysis code.
Azuma, Kisaburo*; Yamaguchi, Yoshihito; Li, Y.
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 7 Pages, 2021/07
Yamaguchi, Yoshihito; Katsuyama, Jinya; Masaki, Koichi*; Li, Y.
JAEA-Research 2020-017, 80 Pages, 2021/02
JAEA-Research-2020-017.pdf:3.5MB
The seismic probabilistic risk assessment (seismic PRA) is an important methodology to evaluate the seismic safety of nuclear power plants. Regarding seismic fragility evaluations performed in the seismic PRA, the Probabilistic Fracture Mechanics (PFM) can be applied as a useful evaluation technique for aged piping with crack or wall thinning due to the age-related degradation. Here, to advance seismic PRA methodology for the long-term operated nuclear power plants, a guideline for the fragility evaluation on the typical aged piping of nuclear power plants has been developed taking the aged-related degradation into account.
Yamaguchi, Yoshihito; Mano, Akihiro; Katsuyama, Jinya; Masaki, Koichi*; Miyamoto, Yuhei*; Li, Y.
JAEA-Data/Code 2020-021, 176 Pages, 2021/02
JAEA-Data-Code-2020-021.pdf:5.26MB
In Japan Atomic Energy Agency, as a part of researches on the structural integrity assessment and seismic safety assessment of aged components in nuclear power plants, a probabilistic fracture mechanics (PFM) analysis code PASCAL-SP (PFM Analysis of Structural Components in Aging LWR - Stress Corrosion Cracking at Welded Joints of Piping) has been developed to evaluate failure probability of piping. The initial version was released in 2010, and after that, the evaluation targets have been expanded and analysis functions have been improved based on the state-of-the art technology. Now, it is released as Ver. 2.0. In the latest version, primary water stress corrosion cracking in the environment of Pressurized Water Reactor, nickel based alloy stress corrosion cracking in the environment of Boiling Water Reactor, and thermal embrittlement can be taken into account as target age-related degradation. Also, many analysis functions have been improved such as incorporations of the latest stress intensity factor solutions and uncertainty evaluation model of weld residual stress. Moreover, seismic fragility evaluation function has been developed by introducing evaluation methods including crack growth analysis model considering excessive cyclic loading due to large earthquake. Furthermore, confidence level evaluation function has been incorporated by considering the epistemic and aleatory uncertainties related to influence parameters in the probabilistic evaluation. This report provides the user's manual and analysis methodology of PASCAL-SP Ver. 2.0.
Yamaguchi, Yoshihito; Hasegawa, Kunio; Li, Y.
Journal of Pressure Vessel Technology, 142(4), p.041507_1 - 041507_6, 2020/08
Times Cited Count：0 Percentile：0(Engineering, Mechanical)The phenomenon of crack closure is important in the prediction of fatigue crack growth. Several experimental data indicate the closing of fatigue cracks both under negative and positive loads at constant amplitude loading cycles, depending on the magnitude of stress amplitude and stress ratio. Appendix A-4300 of the ASME Code Section XI provides two equations of fatigue crack growth rates expressed by the stress intensity factor range for ferritic steels under negative stress ratio. The boundary of two fatigue crack growth rates is classified with the magnitude of applied stress intensity factor range, in consideration of the crack closure. The boundary value provided by the ASME Code Section XI is validated in this study through an investigation of the influence of the magnitude of the applied stress intensity factor range on crack closure, with the application of fatigue crack growth tests using ferritic steel specimens in air environment at room and high temperatures. Crack closures are obtained as a parameter of stress ratio, and herein, were found to occur at a smaller applied stress intensity factor range, as opposed to the definition given by Appendix A-4300.
Yamaguchi, Yoshihito; Katsuyama, Jinya; Kaji, Yoshiyuki; Osaka, Masahiko; Li, Y.
Mechanical Engineering Journal (Internet), 7(3), p.19-00560_1 - 19-00560_12, 2020/06
Since the Fukushima Daiichi Nuclear Power Plant accident, we have been developing a failure evaluation method that considers creep damage mechanisms using detailed three-dimensional finite element analysis model of lower head including penetration, stub tubes, and weld parts, etc., for the early completion of the decommissioning of the nuclear power plants in Fukushima Daiichi. For the finite element analysis, we have been obtaining material properties for which no data are provided in existing databases or in the literature. In particular, creep data corresponding to the high temperature region near the melting point of materials is important in evaluating creep deformation under severe accident conditions. In this study, we obtained the uniaxial tensile and creep properties for low-alloy steel, stainless steel, and Ni-based alloy. In particular, creep test data with long rupture times at high temperatures are expanded using a tensile test machine that can measure the elongation of test specimens in a noncontact measurement system. The parameters related to the failure evaluation were improved on the basis of the expanded creep database.
Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.; Onizawa, Kunio
Journal of Pressure Vessel Technology, 142(2), p.021906_1 - 021906_11, 2020/04
Times Cited Count：1 Percentile：8.01(Engineering, Mechanical)Mano, Akihiro; Katsuyama, Jinya; Miyamoto, Yuhei*; Yamaguchi, Yoshihito; Li, Y.
International Journal of Pressure Vessels and Piping, 179, p.103945_1 - 103945_6, 2020/01
Times Cited Count：1 Percentile：12.35(Engineering, Multidisciplinary)Weld residual stress (WRS) is one of the most important factors in the structural integrity assessment of piping welds, and it is considered a driving force for crack growth. It is characterized by large uncertainty. For more rational assessment, it is important to consider the uncertainty of WRS for evaluating crack growth behavior in probabilistic fracture mechanics (PFM) analysis. In existing PFM analysis codes, WRS uncertainty is set by statistically processing the results of multiple finite element analyses. This process depends on the individual performing PFM analysis, which may lead to uncertainties whose sources would be different from the original WRS. In this study, we developed a new WRS evaluation model based on Fourier transformation, and the model was incorporated into PASCAL-SP, which has been developed by Japan Atomic Energy Agency. Through improvements to the code, WRS uncertainty can be considered automatically and appropriately by inputting multiple WRS analysis results directly as input data for PFM analysis.
Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.
Journal of Nuclear Engineering and Radiation Science, 5(3), p.031505_1 - 031505_8, 2019/07
Probabilistic fracture mechanics (PFM) analysis is expected as a rational method for the structural integrity assessment because it can consider the uncertainties of various influence factors and can evaluate the quantitative value such as failure probability of a cracked component as the solution. In the Japan Atomic Energy Agency, a PFM analysis code PASCAL-SP has been developed for the structural integrity assessment of piping welds in nuclear power plants. In the latest few decades, a number of cracks due to primary water stress corrosion cracking (PWSCC) have been detected in the nickel-based alloy welds in the primary piping of pressurized water reactors (PWRs). Thus the structural integrity assessment taking account of PWSCC has become important. In this paper, we improved PASCAL-SP for the assessment considering PWSCC by introducing the several analytical functions such as the evaluation models of crack initiation time, crack growth rate and probability of crack detection. By using improved PASCAL-SP, the failure probabilities of pipes with a circumferential crack or an axial crack due to PWSCC were evaluated as numerical examples. We also evaluated the influence of a leak detection and a non-destructive examination on the failure probabilities. On the basis of the numerical results, we concluded that the improved PASCAL-SP is useful for evaluating the failure probability of pipe taking PWSCC into account.
