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Journal Articles

Improvement of probabilistic fracture mechanics analysis code PASCAL-SP regarding stress corrosion cracking in nickel based alloy weld joint of piping system in boiling water reactor

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

Journal of Pressure Vessel Technology, 144(1), p.011506_1 - 011506_9, 2022/02

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.

Journal Articles

A Novel method to uniquely determine the parameters in Gurson-Tvergaard-Needleman model

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:0 Percentile:0(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.

Journal Articles

Benchmarking study on probabilistic fracture mechanics analysis codes xLPR and PASCAL-SP considering primary water stress corrosion cracking

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

Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 7 Pages, 2021/07

Probabilistic fracture mechanics (PFM) is expected as a more rational methodology for the structural integrity assessments of nuclear power components because it can consider the inherent probabilistic distributions of various influencing factors and quantitatively evaluate the failure probabilities of the components. The Japan Atomic Energy Agency (JAEA) has developed a PFM analysis code, PASCAL-SP, to evaluate the failure probabilities of piping caused by aging degradation mechanisms, such as fatigue and stress corrosion cracking in the environments of both pressurized water and boiling water reactors. To improve confidence in the analysis results obtained from PASCAL-SP, a benchmarking study was conducted together with the PFM analysis code, xLPR, which was developed jointly by the U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute. The benchmarking study was composed of deterministic and probabilistic analyses related to primary water stress corrosion cracking in a dissimilar metal weld joint in a pressurized water reactor surge line. The analyses were conducted independently by NRC staff and JAEA using their own codes and under common analysis conditions. In the present paper, the analysis conditions for the deterministic and probabilistic analyses are described in detail, and the analysis results obtained from the xLPR and PASCAL-SP codes are presented. It was confirmed that the analysis results obtained from the two codes were in good agreement.

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

Probabilistic fracture mechanics benchmarking study involving the xLPR and PASCAL-SP codes; Analysis by PASCAL-SP

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

Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 7 Pages, 2020/08

A probabilistic fracture mechanics (PFM) analysis code PASCAL-SP has been developed by Japan Atomic Energy Agency (JAEA) for evaluating the failure probability of piping in nuclear power plant considering aged-related degradations such as primary water stress corrosion cracking (PWSCC) in pressurized water reactor environments and fatigue. To strengthen the confidence of analysis results, benchmarking study is being performed with PFM analysis code xLPR which has been developed by U.S.NRC in collaboration with EPRI. The benchmarking study consists of deterministic and probabilistic analyses on PWSCC under the common analysis conditions. In addition, deterministic sensitivity analysis on weld residual stress distributions is also included in the benchmarking study. These analyses are carried out by U.S.NRC and JAEA independently using their own codes. At current stage, the deterministic analyses by both xLPR and PASCAL-SP codes have been finished and probabilistic analyses are underway. This paper presents the details of conditions and comparisons of the results between the two codes in the deterministic analyses. In the deterministic analyses, both codes provided almost the same results including the values of stress intensity factor. In addition, probabilistic analysis conditions and results obtained from PASCAL-SP are presented.

Journal Articles

Improved Bayesian update method on flaw distributions reflecting non-destructive inspection result

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

Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 8 Pages, 2020/08

We have developed a probabilistic fracture mechanics (PFM) analysis code PASCAL4 for evaluating failure frequency of reactor pressure vessels (RPVs). It is known that flaw distributions have an important role in failure frequency calculation in PFM analysis. Previously, we proposed likelihood function to obtain more realistic flaw distributions applicable for both case when flaws are detected and when there is no flaw indication as the inspection results based on Bayesian update methodology. Here, it can be applied to independently obtain posterior distributions of flaw depth and density. In this study, we improve the likelihood function to enable them to update flaw depth and density simultaneously. Based on the improved likelihood function, an example is presented in which flaw distributions are estimated by reflecting NDI results through Bayesian update and PFM analysis. The results indicate that the improved likelihood functions are useful for estimating flaw distributions.

Journal Articles

Influence evaluation of sampling methods of the nondestructive examination on failure probability of piping based on probabilistic fracture mechanics analysis

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

Mechanical Engineering Journal (Internet), 7(3), p.19-00567_1 - 19-00567_11, 2020/06

Non-destructive examinations (NDEs) have an important role in assurance of the structural integrity of nuclear components including pipe lines. In Japanese nuclear power plants, NDEs are performed for welds in piping in accordance with the rules such as the Rules on Fitness-for-Service for Nuclear Power Plants of the Japan Society of Mechanical Engineers. For the welds where stress corrosion cracking (SCC) is not postulated, NDEs are performed in each 10-year interval. For each interval, the extent of examination is specified in the rules. In general, there are two kinds of sampling method for selecting welds to be examined in each interval considering the specified extent of examination. The first method is the fixed location sampling method, in which welds for NDEs are same as those examined in the last interval. The second method is the random location sampling method, in which welds for NDEs are selected from those not examined in the last interval. The selection of the sampling method is important to assure the structural integrity of piping. Probabilistic fracture mechanics (PFM) analysis which is one of rational structural integrity assessment methods can quantitatively calculate failure probability of welds in piping considering aging degradation mechanisms such as SCC and fatigue as well as crack detections and repair of cracked welds through NDE. In this study, to clarify the influence of the sampling methods on structural integrity of piping, we evaluated the failure probability of a typical nuclear piping considering NDEs based on the two sampling methods through PFM analysis. From the results, we clarified the quantitative influence of two sampling methods on failure probability of piping.

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:0(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

Comparison of two-phase critical flow models for estimation of leak flow rate through cracks

Watanabe, Tadashi*; Katsuyama, Jinya; Mano, Akihiro

International Journal of Nuclear and Quantum Engineering (Internet), 13(11), p.516 - 519, 2019/10

The estimation of leak flow rates through narrow cracks in structures is of importance for nuclear reactor safety, since the leak flow could be detected before occurrence of loss-of-coolant accidents. The two-phase critical leak flow rates are calculated using the system analysis code, and two representative non-homogeneous critical flow models, Henry-Fauske model and Ransom-Trapp model, are compared. The pressure decrease and vapor generation in the crack, and the leak flow rates are found to be larger for the Henry-Fauske model. It is shown that the leak flow rates are not affected by the structural temperature, but affected largely by the roughness of crack surface.

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

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

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

 Times Cited Count:7 Percentile:56.85(Engineering, Mechanical)

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

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

Safety design approach for external events in Japan Sodium-cooled Fast Reactor

Yamano, Hidemasa; Kubo, Shigenobu; Tani, Akihiro*; Nishino, Hiroyuki; Sakai, Takaaki

Proceedings of 2012 International Congress on Advances in Nuclear Power Plants (ICAPP '12) (CD-ROM), p.813 - 825, 2012/06

This paper describes a safety design approach for external events in the design study of Japan Sodium-cooled Fast Reactor, in parallel to the approach for internal events which is conformed to the defence-in-depth principle in IAEA. An emphasis is the addition of new eventcategory that has to be considered in design, i.e., introduction of "design extension condition (DEC)." In this study, the DEC category was extended to the external events. In addition to seismic design, other external events such as tsunami, strong wind and abnormal temperature were addressed in this study. From a wide variety of external events consisting of natural hazards and human-induced ones, a screening method was developed in terms of sitting, consequence, frequency to select representative events. Design approaches for these events were categorized on the probabilistic, statistical and deterministic basis.

Journal Articles

Conceptual design study for the demonstration reactor of JSFR, 3; Safety design and evaluation

Tani, Akihiro*; Shimakawa, Yoshio*; Kubo, Shigenobu*; Fujimura, Ken; Yamano, Hidemasa

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10

JAEA Reports

Performance test results of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system (Contract research)

Inagaki, Yoshiyuki; Hayashi, Koji; Kato, Michio; Fujisaki, Katsuo; Aita, Hideki; Takeda, Tetsuaki; Nishihara, Tetsuo; Inaba, Yoshitomo; Ohashi, Hirofumi; Katanishi, Shoji; et al.

JAERI-Tech 2003-034, 129 Pages, 2003/05

JAERI-Tech-2003-034.pdf:7.62MB

no abstracts in English

Oral presentation

Definition of external event conditions for safety design in Japan Sodium-cooled Fast Reactor

Yamano, Hidemasa; Kubo, Shigenobu; Nishino, Hiroyuki; Sakai, Takaaki; Tani, Akihiro*

no journal, , 

The importance in safety measures against external events becomes more significant recently and investigation for the design measures is inevitable. Thus conditions for external event were established in order to investigate safety measures for JSFR. In this paper, we report a screening method and definition method for postulating condition for the design.

Oral presentation

A New probabilistic evaluation model on weld residual stress

Katsuyama, Jinya; Miyamoto, Yuhei*; Yamaguchi, Yoshihito; Mano, Akihiro; Li, Y.

no journal, , 

Weld residual stress (WRS) is one of the most important factors with a great deal of uncertainty, which is considered as a driving force for crack growth in the structural integrity assessment of piping welds. For more rational assessments, it is important to consider the uncertainty of WRS in probabilistic fracture mechanics (PFM) analysis. In the existing PFM analysis codes, the uncertainty of WRS is set through statistical process of multiple finite element analysis (FEA) results. This process depends on persons who perform PFM analysis, and it may give different uncertainties. In this study, we developed a new WRS evaluation model based on the Fourier transformation, and the model was introduced into PASCAL-SP which has been developed by Japan Atomic Energy Agency. Through these improvements of the code, the uncertainty of WRS can be taken into account automatically and appropriately by inputting multiple WRS analysis results directly as input data of PFM analysis.

29 (Records 1-20 displayed on this page)