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ASME 2010 Pressure Vessels and Piping Conference: Volume 6, Parts A and B | 2010

Investigation on Evaluation Method Based on J Integral for Retardation of Crack Growth Due to Excessive Loading Beyond Small Scale Yielding Condition

Yoshihito Yamaguchi; Jinya Katsuyama; Kunio Onizawa; Hideharu Sugino; Yinsheng Li

Niigata-ken Chuetsu-Oki earthquake occurred in July 2007, whose magnitude was beyond the assumed one provided in seismic design. Therefore it becomes an important issue to evaluate the effect of excessive loading, in particular, for the components with existing crack. Fatigue crack growth rate is usually expressed by Paris’s law using the range of stress intensity factor (ΔK). However, applicability of the model to loading conditions beyond the small scale yielding remains as an issue since ΔK is inappropriate in such a high loading level. In this study, the fatigue crack growth behaviors after applying the excessive loads were investigated using austenitic stainless steel and carbon steel. Instead of ΔK, J-integral value for crack growth evaluation due to cyclic loading has been applied based on the experimental data to treat the excessive loading beyond small scale yielding. The finite element method (FEM) analyses were conducted to evaluate the stress distribution and plastic zone size for the excessive loading condition. The modified Wheeler model using J-integral range, ΔJ, has been proposed for the prediction of retardation effect on crack growth after excessive loading. It was indicated that retardation effect by excessive loading beyond small-scale yielding could be quantitatively evaluated using the J-Wheeler model.Copyright


ASME 2015 Pressure Vessels and Piping Conference | 2015

Development of J-Integral Solutions for Semi-Elliptical Circumferential Cracked Pipes Subjected to Internal Pressure and Bending Moment

Makoto Udagawa; Jinya Katsuyama; Yoshihito Yamaguchi; Yinsheng Li; Kunio Onizawa

The J-integral solutions for cracked pipes are important in crack growth calculation and failure evaluation based on the elastic-plastic fracture mechanics. One of the most important crack types in structural integrity assessment for nuclear piping systems is circumferential semi-elliptical surface crack on the inside of the pipes. Although several J-integral solutions have been provided, no solutions were developed at both the deepest and the surface points of circumferential semi-elliptical surface cracks in pipes. In this study, with backgrounds described above, the J-integral solutions of circumferential semi-elliptical surface cracks on the inside of the pipe were developed by numerical finite element analyses. Three dimensional elastic-plastic analyses were performed considering different material properties, pipe sizes, crack dimensions and, especially, combined loading condition of internal pressure and bending moment which is a typical loading condition for nuclear piping systems. The J values at both the deepest and the surface points were extracted from finite element analysis results. Moreover, in order to benefit users in practical applications, a pair of convenient J-integral estimation equations were developed based on the calculated J values at the deepest and the surface points. Finally, the accuracy and applicability of the convenient equations were confirmed by comparing with the provided stress intensity factor solutions in elastic region and with finite element analysis results in elastic-plastic region.Copyright


ASME 2015 Pressure Vessels and Piping Conference | 2015

Crack Growth Evaluation for Cracked Carbon and Stainless Steel Pipes Under Large Seismic Cyclic Loading

Yoshihito Yamaguchi; Jinya Katsuyama; Yinsheng Li; Kunio Onizawa

Japanese nuclear power plants have recently experienced several large earthquakes beyond the previous design basis ground motion. In addition, cracks resulting from long-term operation have been detected in piping lines. Therefore, it is very important to establish a crack growth evaluation method for cracked pipes that are subjected to large seismic cyclic response loading. In our previous study, we proposed an evaluation method for crack growth during large earthquakes through experimental study using small specimens. In the present study, crack growth tests were conducted on pipes with a circumferential through-wall crack, considering large seismic cyclic response loading with complex wave forms. The predicted crack growth values are in good agreement with the experimental results for both stainless and carbon steel pipe specimens and the applicability of the proposed method was confirmed.Copyright


Volume 1: Plant Operations, Maintenance, Engineering, Modifications, Life Cycle and Balance of Plant; Nuclear Fuel and Materials; Plant Systems, Structures and Components; Codes, Standards, Licensing and Regulatory Issues | 2014

Improvement of Probabilistic Fracture Mechanics Analysis Code for Reactor Piping Considering Large Earthquakes

Yoshihito Yamaguchi; Jinya Katsuyama; Makoto Udagawa; Kunio Onizawa; Yutaka Nishiyama; Yinsheng Li

The probabilistic fracture mechanics analysis code PASCAL-SP is improved by introducing crack-growth evaluation methods based on J-integrals, including calculation functions of J-integral values for semi-elliptical surfaces and through-wall cracks in pipes. Using the improved PASCAL-SP, sensitivity analyses that varied parameters such as earthquake magnitude were carried out on the basis of probabilistic evaluation. Results obtained from sensitivity analyses are also presented, e.g., the effect of earthquake magnitude on failure probability. The improved PASCAL-SP makes evaluation of the failure probability of piping under large seismic loading possible.Copyright


Journal of Nuclear Materials | 2014

Effects of thermal aging on microstructure and hardness of stainless steel weld-overlay claddings of nuclear reactor pressure vessels

T. Takeuchi; Y. Kakubo; Y. Matsukawa; Y. Nozawa; T. Toyama; Yasuyoshi Nagai; Yutaka Nishiyama; J. Katsuyama; Yoshihito Yamaguchi; Kunio Onizawa; Masahide Suzuki


Journal of Nuclear Materials | 2014

Effects of neutron irradiation on microstructures and hardness of stainless steel weld-overlay cladding of nuclear reactor pressure vessels

T. Takeuchi; Y. Kakubo; Y. Matsukawa; Y. Nozawa; T. Toyama; Yasuyoshi Nagai; Yutaka Nishiyama; J. Katsuyama; Yoshihito Yamaguchi; Kunio Onizawa


Nuclear Engineering and Design | 2014

Effect of cyclic loading on the relaxation of residual stress in the butt-weld joints of nuclear reactor piping

Jinya Katsuyama; Yoshihito Yamaguchi; Yinsheng Li; Kunio Onizawa


Journal of The Society of Materials Science, Japan | 2010

Bending Fatigue Strength of Austenitic Stainless Steel SUS316 in Mercury

Takashi Naoe; Yoshihito Yamaguchi; Masatoshi Futakawa; Takashi Wakui


Mechanical Engineering Journal | 2016

Development of failure evaluation method for BWR Lower head in severe accident; - Creep damage evaluation based on thermal-hydraulics and structural analyses -

Jinya Katsuyama; Yoshihito Yamaguchi; Yoshiyuki Nemoto; Yoshiyuki Kaji; Hiroyuki Yoshida


Journal of Nuclear Materials | 2012

Quantification of fatigue crack propagation of an austenitic stainless steel in mercury embrittlement

Takashi Naoe; Yoshihito Yamaguchi; Masatoshi Futakawa

Collaboration


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Jinya Katsuyama

Japan Atomic Energy Agency

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Yinsheng Li

Japan Atomic Energy Agency

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Kunio Onizawa

Japan Atomic Energy Agency

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Takashi Naoe

Japan Atomic Energy Agency

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Yutaka Nishiyama

Japan Atomic Energy Research Institute

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Hiroyuki Yoshida

Japan Atomic Energy Agency

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Yoshiyuki Kaji

Japan Atomic Energy Agency

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Makoto Udagawa

Japan Atomic Energy Agency

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Hiroyuki Kogawa

Japan Atomic Energy Agency

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