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Journal Articles

Influence evaluation of sampling methods of the non-destructive examination on failure probability of piping based on probabilistic fracture mechanics analyses

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

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

In Japanese nuclear power plants, non-destructive examinations (NDEs) are performed for welds in piping in accordance with the rules such as Rules on Fitness-for-Service for Nuclear Power Plant of the Japan Society of Mechanical Engineers (JSME FFS). A set of NDEs is performed in each 10-year interval, and the extent of examination in each interval is specified in the rules. Welding lines to be examined are selected considering the extent of examination based on two sampling methods. One is the fixed location sampling method that welds to be examined are selected from welds examined in the last interval. The other is the random location sampling method that welds to be examined are selected from other than welds examined in the last interval. The selection of the sampling methods is considered to be one of the important factors in in-service inspection. Probabilistic fracture mechanics (PFM) analysis is expected to be more rational method for the structural integrity assessment because it can consider the uncertainties of various influence factors and evaluate the quantitative values such as failure probability of a cracked component as the solution. In this study, we investigated the influence of the sampling methods related to the NDE on failure probability of typical nuclear piping based on PFM analyses. Through sensitivity PFM analyses, we confirmed that failure probability value obtained from PFM analysis is useful as a quantitative numerical index for selecting the sampling method in an in-service inspection.

Journal Articles

Verification of a probabilistic fracture mechanics code PASCAL4 for reactor pressure vessels

Lu, K.; Katsuyama, Jinya; Li, Y.; Miyamoto, Yuhei*; Hirota, Takatoshi*; Itabashi, Yu*; Nagai, Masaki*; Suzuki, Masahide*; Kanto, Yasuhiro*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05

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

Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses

Katsuyama, Jinya; Uno, Shumpei*; Watanabe, Tadashi*; Li, Y.

Frontiers of Mechanical Engineering, 13(4), p.563 - 570, 2018/12

For the structural integrity assessments on reactor pressure vessels (RPVs) under pressurized thermal shock (PTS) events, thermal hydraulic (TH) behavior of coolant water is one of the most important influence factors. Configuration of plant equipment and their dimensions, and operator action have large influences on TH behavior. In this study, to investigate the influence of operator action on TH behavior during a PTS event, we developed an analysis model for a typical Japanese plant, and performed TH and structural analyses. Two different operator action times were assumed based on the Japanese and US' rules. From the analysis results, it was clarified that differences in operator action times have a significant effect on TH behavior and loading conditions, that is, following the Japanese rule may lead to lower stresses compared to that when following the US rule because earlier operator action caused lower pressure in the RPV.

Journal Articles

Analyses of LSTF experiment and PWR plant for 5% cold-leg break loss of coolant accident

Watanabe, Tadashi*; Ishigaki, Masahiro*; Katsuyama, Jinya

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 9 Pages, 2018/10

The analyses of LSTF experiment and PWR plant for 5% cold-leg break LOCA are performed using the RELAP5/MOD3.3 code. The discharge coefficient of critical flow model is determined so as to obtain the agreement of pressure transient between the LSTF experiment and the experimental analysis, and used for the PWR analysis. The characteristics of thermal-hydraulic phenomena in the experiment are shown to be simulated well by the two analyses. The decrease in core differential pressure during the loop-seal clearing is, however, underestimated by the two analyses, and the core heat up is not predicted. The loop flow rates are also underestimated by the two analyses. Although the duration of core heat up during the boil-off period is longer in the experimental analysis, the results of two analyses agree well, and the effect of scaling is found to be small between the experimental analysis and the PWR analysis.

Journal Articles

Development of crack evaluation models for probabilistic fracture mechanics analyses of Japanese reactor pressure vessels

Lu, K.; Masaki, Koichi; Katsuyama, Jinya; Li, Y.

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

Journal Articles

Development of probabilistic fracture mechanics analysis code PASCAL Version 4 for reactor pressure vessels

Lu, K.; Masaki, Koichi; Katsuyama, Jinya; Li, Y.; Uno, Shumpei*

Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 10 Pages, 2018/07

Journal Articles

Verification of probabilistic fracture mechanics analysis code through benchmark analyses

Li, Y.; Uno, Shumpei*; Masaki, Koichi; Katsuyama, Jinya; Dickson, T.*; Kirk, M.*

Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 11 Pages, 2018/07

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

Sensitivity study on the effects of nondestructive examinations on failure probabilities of reactor pressure vessels

Arai, Kensaku*; Katsuyama, Jinya; Li, Y.

Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 7 Pages, 2018/07

We have developed an analysis code called PASCAL for assessing failure frequencies of RPVs based on probabilistic fracture mechanics. In failure frequency analyses, flaw size distribution in RPVs is one of the most important parameters, and it is determined by considering possible flaws generated during fabrication and the flaw-detection capabilities of nondestructive examinations (NDEs). In this study, the effects of NDEs on failure frequencies of RPVs are evaluated using PASCAL considering probability of detection (POD) in terms of minimum detectable flaw size, the smallest probability of non-detection (PND), and flaw size where POD value reaches the smallest PND.

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

Development of stress intensity factors for subsurface flaws in plates subjected to polynomial stress distributions

Lu, K.; Mano, Akihiro; Katsuyama, Jinya; Li, Y.; Iwamatsu, Fuminori*

Journal of Pressure Vessel Technology, 140(3), p.031201_1 - 031201_11, 2018/06

 Percentile:100(Engineering, Mechanical)

JAEA Reports

User's manual and analysis methodology of probabilistic fracture mechanics analysis code PASCAL Ver.4 for reactor pressure vessel (Contract research)

Katsuyama, Jinya; Masaki, Koichi; Miyamoto, Yuhei*; Li, Y.

JAEA-Data/Code 2017-015, 229 Pages, 2018/03

JAEA-Data-Code-2017-015.pdf:5.8MB
JAEA-Data-Code-2017-015(errata).pdf:0.15MB

As a part of the structural integrity research for aging light water reactor components, a probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed in JAEA. The PASCAL code can evaluate the conditional failure probabilities and failure frequencies for core region in reactor pressure vessels under the pressurized thermal shock events. In this study, we improved many functions such as the stress intensity factor solutions, the fracture toughness models, or confidence level evaluation function by considering epistemic and aleatory uncertainties related to influence parameters in the structural integrity assessment. We also developed the analysis module PASCAL-Manager which calculates the failure frequency for the entire core region taking into consideration the failure probabilities obtained from PACAL-RV. Based on these improvements, the new analysis code is upgraded to PASCAL Ver.4. This report provides the user's manual and theoretical background of PASCAL Ver.4.

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

An Estimation method of flaw distributions reflecting inspection results through Bayesian update

Lu, K.; Miyamoto, Yuhei*; Mano, Akihiro; Katsuyama, Jinya; Li, Y.

Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 9 Pages, 2017/11

Nowadays, probabilistic fracture mechanics (PFM) has been utilized in several countries as a rational method for structural integrity assessment of important structural components such as reactor pressure vessels (RPVs). In PFM analyses, potential flaws in target components are used to evaluate the failure probability or frequency. Therefore, flaw distributions (i.e., flaw depth and density distributions) in an RPV shall be rationally set as one of the most important influential factors, which are developed during the manufacturing process such as welding. Recently, a Bayesian updating methodology was applied to reflect the inspection results into flaw distributions, and the likelihood functions applicable to the case when flaws are detected in inspections were proposed. However, there may be no flaw indication as the inspection results of some RPVs. The flaw distributions in this situation are important while the corresponding likelihood functions have not been proposed. Therefore, this study proposed likelihood functions to be applicable for both case when flaws are detected and when there is no flaw indication as the inspection results. Based on the proposed likelihood functions, several application examples were given in which flaw distributions were estimated by reflecting the inspection results through Bayesian update. The results indicate that the proposed likelihood functions are useful for estimating the flaw distribution for the case when there is no flaw indication as the inspection results.

Journal Articles

Benchmark analyses using probabilistic fracture mechanics analysis codes for reactor pressure vessels

Arai, Kensaku*; Katsuyama, Jinya; Li, Y.

Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 8 Pages, 2017/11

Probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed to assess structural integrity of aged reactor pressure vessels (RPVs) of light water nuclear power plants by Japan Atomic Energy Agency (JAEA). PASCAL is able to obtain failure frequency such as through-wall cracking frequency (TWCF) of RPVs under several transients including pressurized thermal shock (PTS) event. On the other hand, FAVOR was developed to perform almost the same analysis by Oak Ridge National Laboratory (ORNL) under United States Nuclear Regulatory Commission (USNRC) funding and has been utilized in the US nuclear regulation. To improve the reliability of PFM analysis results of PASCAL, benchmark analyses between PASCAL and FAVOR were performed. This paper provides results of the benchmark analyses using analysis conditions and parameters of the US 3-loop pressurized water reactor (PWR) nuclear power plant. Furthermore, sensitivity analyses relating to differences of analysis models (ex. Embrittlement correlation model) between Japan and the US were also conducted.

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

Verification of probabilistic fracture mechanics analysis code PASCAL through benchmark analyses with FAVOR

Li, Y.; Uno, Shumpei*; Katsuyama, Jinya; Dickson, T.*; Kirk, M.*

Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 8 Pages, 2017/07

Journal Articles

Probabilistic fracture mechanics analysis models for Japanese reactor pressure vessels

Lu, K.; Katsuyama, Jinya; Uno, Shumpei; Li, Y.

Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 8 Pages, 2017/07

206 (Records 1-20 displayed on this page)