Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 77

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

JAEA Reports

Guideline on seismic fragility evaluation for aged piping (Contract research)

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.

JAEA Reports

User's manual and analysis methodology of probabilistic fracture mechanics analysis code PASCAL-SP Ver. 2 for piping (Contract research)

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.

Journal Articles

Non-invasive imaging of radiocesium dynamics in a living animal using a positron-emitting $$^{127}$$Cs tracer

Suzui, Nobuo*; Shibata, Takuya; Yin, Y.-G.*; Funaki, Yoshihito*; Kurita, Keisuke; Hoshina, Hiroyuki*; Yamaguchi, Mitsutaka*; Fujimaki, Shu*; Seko, Noriaki*; Watabe, Hiroshi*; et al.

Scientific Reports (Internet), 10, p.16155_1 - 16155_9, 2020/10

 Times Cited Count:0 Percentile:100(Multidisciplinary Sciences)

Journal Articles

Fatigue crack growth for ferritic steel under negative stress ratio

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:100(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.

Journal Articles

Expansion of high temperature creep test data for failure evaluation of BWR lower head in severe accident

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.

Journal Articles

Neutron emission spectrum from gold excited with 16.6 MeV linearly polarized monoenergetic photons

Kirihara, Yoichi; Nakashima, Hiroshi; Sanami, Toshiya*; Namito, Yoshihito*; Itoga, Toshiro*; Miyamoto, Shuji*; Takemoto, Akinori*; Yamaguchi, Masashi*; Asano, Yoshihiro*

Journal of Nuclear Science and Technology, 57(4), p.444 - 456, 2020/04

 Times Cited Count:1 Percentile:68.44(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Crack growth evaluation for cracked stainless and carbon steel pipes under large seismic cyclic loading

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:100(Engineering, Mechanical)

Journal Articles

A New probabilistic evaluation model for weld residual stress

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:0 Percentile:100(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.

Journal Articles

Improvement of probabilistic fracture mechanics analysis code PASCAL-SP with regard to PWSCC

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.

Journal Articles

Expansion of high temperature creep test data for failure evaluation of BWR lower head in a severe accident

Yamaguchi, Yoshihito; Katsuyama, Jinya; Kaji, Yoshiyuki; Osaka, Masahiko; Li, Y.

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05

After the Fukushima Daiichi (1F) Nuclear Power Plant accident, we have been developing a prediction method for rupture time and location considering creep damage mechanisms using finite element analysis for early completion of the decommissioning of nuclear power plants in 1F. We have also been obtaining material properties at high temperature near the melting point which are not provided in existing database or literature for the finite element analysis. In this study, we performed uni-axial tensile and creep tests for low alloy steel, Ni-based alloy steel and stainless steels and expanded existing database of material properties. Especially, creep data with longer rupture time at high temperature were obtained by a creep test equipment with a noncontact measurement system. To improve the accuracy of failure evaluation under severe accident conditions, we determined parameters of creep constitutive law based on the expanded database.

Journal Articles

Crack growth prediction for cracked dissimilar metal weld joint in pipe under large seismic cyclic loading

Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.

Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 8 Pages, 2018/07

In order to assess the structure integrity of cracked pipes considering occurrence of large earthquakes, crack growth evaluation method for cracked pipes is important. In present study, evaluation method of crack growth by seismic loading was proposed for a dissimilar metal weld joint of nickel based alloy through experimental study using small specimens. Then, validation of the proposed method was performed through crack growth tests by using dissimilar metal weld pipe with circumferential through-wall crack. The predicted crack growth values were in good agreement with the experimental results and the applicability of the proposed method was confirmed.

Journal Articles

Fatigue crack growth for ferritic steel under negative stress ratio

Yamaguchi, Yoshihito; Hasegawa, Kunio; Li, Y.

Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 6 Pages, 2018/07

Crack closure during fatigue crack growth is an important phenomenon for predicting fatigue crack growth amount. Much experimental data shows that fatigue cracks close at not only negative loads but also positive loads during a constant amplitude loading cycle. The Appendix A-4300 in the ASME Code Section XI provides two equations of fatigue crack growth rates expressed by stress intensity factor range for ferritic steels under negative stress ratio. One is the equation taking into account crack closure and the other does not consider the crack closure. The boundary of crack closure is classified by the magnitude of applied stress intensity factor range. The objective of this paper is to investigate the influence of the magnitude of stress intensity factor range on crack closure. Fatigue tests have been performed on ferritic steel in air environment at room and high temperatures. Crack closures were obtained as a parameter of stress ratio. It was found that crack closure occurs more small applied stress intensity factor range than the definition given by the Appendix A-4300.

Journal Articles

Creep deformation analysis of a pipe specimen based on creep damage evaluation method

Katsuyama, Jinya; Yamaguchi, Yoshihito; Li, Y.

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 6 Pages, 2018/07

It has become more important to develop methods for evaluating failure behavior of the nuclear components under severe conditions. We are researching on prediction methods of creep deformation and failure behavior of the nuclear components under elevated temperature conditions based on finite element analysis. In this study, as a part of a project called COSSAL, we performed failure analysis of a large scale pipe experiment to validate our prediction methods based on a creep damage evaluation method. We conclude that creep constitutive law that consider material damage can provide the highest accurate analysis.

Journal Articles

Investigation of crack growth evaluation method under seismic loading by considering effects of load history

Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.

Yosetsu Kozo Shimpojiumu 2017 Koen Rombunshu, p.21 - 27, 2017/12

no abstracts in English

Journal Articles

An Application of the probabilistic fracture mechanics code PASCAL-SP to risk informed in-service inspection for piping

Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.

Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 12 Pages, 2017/11

As a rational inspection methodology, risk informed in-service inspection (RI-ISI) has been widely utilized in in-service inspections of nuclear power plants (NPPs) in several countries. In some of NPPs, an RI-ISI methodology developed by Westinghouse Owners Group (WOG) was applied. As a part of RI-ISI process, extent of examination for important piping segments are determined through the comparisons of leak frequencies with its target value based on the industrial piping leak experiences. The leak frequencies for segments are used as a numerical factor for planning examination based on WOG methodology, and can be evaluated through analyses on the basis of probabilistic fracture mechanics (PFM). In Japan Atomic Energy Agency (JAEA), we have developed a PFM analysis code PASCAL-SP for evaluating leak and rupture probabilities or frequencies of welds in piping of light water reactors taking crack initiation and propagation due to aging degradation mechanisms such as fatigue into consideration. Also, evaluation models of probability of crack detection by non-destructive examination considering the crack type, crack depth and performance of examination team is incorporated in PASCAL-SP. In this study, we investigated the applicability of PASCAL-SP into planning of examination considering the effects of repair methodology, performance of inspection team, and examination time. On the basis of analysis results, it was found that examination plans can be reasonably determined by using PASCAL-SP under several conditions, and it was concluded that the PFM is very effective tools in RI-ISI.

Journal Articles

Creep damage evaluations for BWR lower head in severe accident

Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Osaka, Masahiko

Transactions of 24th International Conference on Structural Mechanics in Reactor Technology (SMiRT-24) (USB Flash Drive), 11 Pages, 2017/08

It is difficult to assess rupture behavior of the lower head of reactor pressure vessel in boiling-water-type nuclear power plants due to severe accident like Fukushima Daiichi because Boiling Water Reactor (BWR) lower heads have geometrically complicated structure with a lot of penetrations. Therefore, we have been developing an analysis method to predict time and location of RPV lower head rupture of BWRs considering creep damage mechanisms based on coupled analysis of three-dimensional Thermal-Hydraulics (TH) and thermal-elastic-plastic-creep analyses. In this study, we performed creep damage evaluations to investigate the effects of the debris depth and heat generation locations on failure behavior of lower head. From the analysis results, we discussed the outflow paths of the relocated molten core to the containment, and it was concluded that failure regions of BWR lower head are only the control rod guide tubes or stub tubes under simulated conditions.

Journal Articles

Improvement of probabilistic fracture mechanics analysis code PASCAL-SP with regard to primary water stress corrosion cracking

Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07

Recently, cracks due to primary water stress corrosion cracking (PWSCC) have been detected in nickel based alloy welds in the primary piping of pressurized water reactors. Structural integrity assessments taking PWSCC into account have become important. Probabilistic fracture mechanics (PFM) is expected as one of rational methods for the assessments because it can account for uncertainty of the influencing factors and evaluate the failure probabilities of components. In JAEA, a PFM analysis code PASCAL-SP was developed to evaluate the failure probability of nuclear pipe. This paper details improvement of the PASCAL-SP to evaluate the failure probability taking PWSCC into account. As numerical examples, the failure probabilities for circumferential and axial cracks due to PWSCC are evaluated. Influence of inspection on failure probabilities are evaluated. As the results, we conclude that the improved PASCAL-SP is useful for evaluating the failure probability taking PWSCC into account.

Journal Articles

Reference curve of fatigue crack growth for ferritic steels under negative R ratio provided by ASME Code Section XI

Hasegawa, Kunio*; Mares, V.*; Yamaguchi, Yoshihito

Journal of Pressure Vessel Technology, 139(3), p.034501_1 - 034501_5, 2017/06

 Times Cited Count:2 Percentile:78.16(Engineering, Mechanical)

Journal Articles

Failure evaluation analysis of reactor pressure vessel lower head in BWR due to severe accident

Kaji, Yoshiyuki; Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Osaka, Masahiko

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04

To investigate the inhomogeneous temperature and stress distribution by geometrical complex of BWR lower head, the detailed 3D model of RPV lower head with control rod guide tubes and shroud supports are constructed and the 3D thermal hydraulic analysis of simulated molten pool and creep deformation analysis of lower head are performed using ANSYS Fluent / Mechanical finite element code. It is found that failure for BWR lower head might be caused by combination between melting failure in inner surface of lower head and creep failure in outer surface of lower head.

Journal Articles

Failure probability analysis of aged piping using probabilistic fracture mechanics methodology considering seismic loads

Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.

Proceedings of 2016 ASME Pressure Vessels and Piping Conference (PVP 2016) (Internet), 10 Pages, 2016/07

77 (Records 1-20 displayed on this page)