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Journal Articles

Proposal of a new subsurface-to-surface flaw transformation rule for fatigue crack growth analyses

Lacroix, V.*; Bouydo, A.*; Katsumata, Genshichiro*; Li, Y.; Hasegawa, Kunio

Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 7 Pages, 2017/07

Journal Articles

Introduction of subsurface proximity criteria in the world and stress intensity factors for transformed surface flaws

Hasegawa, Kunio; Li, Y.; Katsumata, Genshichiro*; Dulieu, P.*; Lacroix, V.*

Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 6 Pages, 2017/07

Net-section stress at the ligament between component free surface and subsurface flaw increases when the ligament distance is short. It can be easily expected that stress intensity factors increase when the subsurface flaw locates near the free surface. To avoid catastrophic failures caused by ligament failure, fitness-for-service (FFS) codes provide flaw-to-surface proximity rules. The proximity rules are used to determine whether the flaws should be treated as subsurface flaws as-is, or transformed to surface flaws. The stress intensity factor for the transformed surface flaw increases furthermore. The increment of the stress intensity factor before and after transformation depends on the location of the subsurface flaw. Although the concept of the proximity rules are the same, the specific criteria for the rules on transforming subsurface flaws to surface flaws differ amongst FFS codes. Particularly, the criteria are different amongst the same organizations of ASME (American Society of Mechanical Engineers). The proximity criteria of the FFS codes in the world were introduced in this paper. In addition, the stress intensity factors based on the different criteria used in the ASME Codes are compared.

Journal Articles

Verification of probabilistic fracture mechanics analysis code PASCAL

Li, Y.; Katsumata, Genshichiro*; Masaki, Koichi*; Hayashi, Shotaro*; Itabashi, Yu*; Nagai, Masaki*; Suzuki, Masahide*; Kanto, Yasuhiro*

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07

In Japan, a PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed by the Japan Atomic Energy Agency to evaluate the through-wall cracking frequencies of Japanese reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock transients. In this study, as a part of the verification activities, a working group was established in Japan, with seven organizations from industry, universities and institutes voluntarily participating as members. The source program of PASCAL was released to the members of the working group. Through one year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group including the verification plan, approaches and results.

Journal Articles

Effect of interaction of embedded crack and free surface on remaining fatigue life

Katsumata, Genshichiro*; Lacroix, V.*; Li, Y.

AIMS Materials Science, 3(4), p.1748 - 1758, 2016/12

Journal Articles

Remaining lives of fatigue crack growths for pipes with subsurface flaws and subsurface-to-surface flaw proximity rules

Katsumata, Genshichiro*; Li, Y.; Hasegawa, Kunio*; Lacroix, V.*

Journal of Pressure Vessel Technology, 138(5), p.051402_1 - 051402_5, 2016/10

 Times Cited Count:3 Percentile:72.68(Engineering, Mechanical)

If a subsurface flaw is located near a component surface, the subsurface flaw is transformed to a surface flaw in accordance with a subsurface-to-surface proximity rule. The re-characterization process from subsurface to surface flaw is adopted in all fitness-for-service (FFS) codes. However, the specific criteria of the re-characterizations are different among the FFS codes. Recently, the authors have proposed a new subsurface-to-surface flaw proximity rule based on experimental data and equivalent fatigue crack growth rate calculations. In this study, fatigue crack growth calculations were carried out for pipes with subsurface flaws, using the proximity rule provided in the current codes and the proposed subsurface-to-surface proximity rule. Different pipe sizes, flaw aspect ratios and ligament distances from subsurface flaws to inner surface of pipes were taken into account. As the results, the current proximity rule gives less conservative fatigue lives, when the aspect ratios of the subsurface flaws are small.

Journal Articles

Development of stress intensity factors for surface cracks with large aspect ratio in plates

Li, Y.; Hasegawa, Kunio; Katsumata, Genshichiro; Osakabe, Kazuya*; Okada, Hiroshi*

Journal of Pressure Vessel Technology, 137(5), p.051207_1 - 051207_8, 2015/10

 Times Cited Count:5 Percentile:64.71(Engineering, Mechanical)

A number of surface cracks with large aspect ratio have been detected in components of nuclear power plants in recent years. The depths of these cracks are even larger than the half of crack lengths. However, the solutions of the stress intensity factor were not provided for semi-elliptical surface cracks with large aspect ratio in the current fitness-for-service codes. In this study, in order to conduct integrity assessment for cracked components, the solutions of the stress intensity factor were calculated using finite element analysis for semi-elliptical surface cracks with large aspect ratio in plates. Solutions were provided at both the deepest and the surface points of the surface cracks. Some of solutions were compared with the available existing results. As the result, it was concluded that the solutions proposed in this paper are applicable in engineering applications.

Journal Articles

Development of probabilistic evaluation models of fracture toughness K$$_{Ic}$$ and K$$_{Ia}$$ for Japanese RPV steels

Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio; Osakabe, Kazuya*; Yoshimoto, Kentaro*

Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 9 Pages, 2015/07

Probabilistic fracture mechanics (PFM) analysis code PASCAL3 has been developed to apply the PFM analysis to the structural integrity assessment of domestic RPVs. In this paper, probabilistic evaluation models of fracture toughness KIc and KIa which have the largest scatter among the associated factors based on the database of Japanese RPV steels are presented. We developed probabilistic evaluation models for KIc and KIa based on the Weibull and lognormal distributions, respectively. The models are compared with the existing lower bound of fracture toughness in the Japanese code and probabilistic model in USA. As the results, the models established in present work satisfy lower bounds of fracture toughness in the Japanese code. The comparison in the models between present work and US showed significant differences that may have an influence on fracture probability of RPV.

Journal Articles

Study on application of PFM analysis method to Japanese code for RPV integrity assessment under PTS events

Osakabe, Kazuya*; Masaki, Koichi*; Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio; Yoshimura, Shinobu*

Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 8 Pages, 2015/07

A probabilistic fracture mechanics (PFM) analysis method for pressure boundary components is useful to evaluate the structural integrity in a quantitative way. This is because the uncertainties related to influence parameters can be rationally incorporated in PFM analysis. From this viewpoint, the probabilistic approach evaluating through-wall cracking frequencies (TWCFs) of reactor pressure vessels (RPVs) has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. As a study of applying PFM analysis to the integrity assessment of domestic RPVs, JAEA has been preparing input data and analysis models to calculate TWCFs using PFM analysis code PASCAL3. In this paper, activities have been introduced such as preparing input data and models for domestic RPVs, verification of PASCAL3, and formulating guideline on general procedures of PFM analysis for the purpose of utilizing PASCAL3. In addition, TWCFs for a model RPV evaluated by PASCAL3 are presented.

Journal Articles

Fatigue crack growth calculations for pipes considering subsurface to surface flaw proximity rules

Katsumata, Genshichiro; Li, Y.; Hasegawa, Kunio; Lacroix, V.*

Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 6 Pages, 2015/07

If a subsurface flaw is located near a component surface, the subsurface flaw is transformed to a surface flaw in accordance with a subsurface-to-surface flaw proximity rule. The re-characterization process from subsurface to surface flaw is adopted in all fitness-for-service (FFS) codes. However, the specific criteria of the re-characterizations are different among the FFS codes. Recently, the authors have proposed a new subsurface-to-surface flaw proximity rule based on experimental data and equivalent fatigue crack growth rates. In this study, fatigue crack growth calculations were carried out for pipes with subsurface flaws, using the proposed subsurface-to-surface flaw proximity rule and the current proximity rule provided in the current JSME and ASME Section XI. Different pipe sizes, flaw aspect ratios and ligament distances from subsurface flaws to inner surface of pipes were taken into account. As the results, the current proximity rule gives less conservative fatigue lives, when the aspect ratios of the subsurface flaws are small.

Journal Articles

Study on structural integrity assessment of reactor pressure vessel based on three-dimensional thermal-hydraulics and structural analyses

Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio; Watanabe, Tadashi*; Nishiyama, Yutaka

Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 9 Pages, 2014/07

For structural integrity assessment on reactor pressure vessel (RPV) of pressurized water reactor during the pressurized thermal shock (PTS) events, temperature of coolant water and heat transfer coefficient between coolant water and RPV are dominant factors. These values can be determined on the basis of thermal-hydraulics (TH) analysis simulating PTS events. Using these values, structural integrity assessment of RPV is performed by thermal-structural analysis, e.g. loading that affects the crack initiation and propagation is evaluated. In this study, we performed the TH and thermal-structural analyses using three-dimensional model of cold-leg, downcomer and RPV to assess loading conditions during the PTS more accurate. We obtained the loading histories at the reactor core region of RPV where a crack is postulated in the structural integrity assessment. Through the comparison between analysis results and current evaluation method, conservatism of current method will be discussed.

Journal Articles

Estimation of through-wall cracking frequency of RPV under PTS events using PFM analysis method for identifying conservatism included in current Japanese code

Osakabe, Kazuya*; Masaki, Koichi*; Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio

Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 7 Pages, 2014/07

The structural integrity of reactor pressure vessel (RPV) during pressurized thermal shock events is judged to be maintained unless the stress intensity factors at the crack tip is smaller than fracture toughness $$K$$$$_{Ic}$$ based on deterministic approach in the current Japanese code. Application of a probabilistic fracture mechanics (PFM) analysis method for the structural reliability assessment of RPVs has become attractive recently, because uncertainties of several parameters can be incorporated rationally. According to the PFM analysis method in the U.S., through-wall cracking frequencies (TWCFs) are estimated. In this study, in order to identify the conservatism in the current code, PFM analyses on TWCF have been performed for certain model of RPVs. The result shows that the current assumption in JEAC 4206-2007 is conservative as compared with realistic conditions. Effects of variation of PTS transients on crack initiation frequency and TWCF have been also discussed.

Journal Articles

Benchmark analysis on probabilistic fracture mechanics analysis codes concerning multiple cracks and crack initiation in aged piping of nuclear power plants

Li, Y.; Osakabe, Kazuya*; Katsumata, Genshichiro; Katsuyama, Jinya; Onizawa, Kunio; Yoshimura, Shinobu*

Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 8 Pages, 2014/07

Multiple cracks in the same welded joints have been detected in piping systems of nuclear power plants. Therefore, structural integrity assessments considering multiple cracks and crack initiation in aged piping have been important. Probabilistic fracture mechanics (PFM) is a rational methodology in structural integrity assessment of aged piping in nuclear power plants. Two PFM codes, PASCAL-SP and PRAISE-JNES, have been improved or developed in Japan for the structural integrity assessment considering the age related degradation mechanisms of pipes. In this paper, a benchmark analysis was conducted considering multiple cracks and crack initiation, in order to confirm their reliability and applicability. Based on the numerical investigation in consideration of important influence factors such as crack number, crack location, crack distribution and crack detection probability of in-service inspection, it was concluded that the analysis results of these two codes are in good agreement.

Oral presentation

Study of system safety evaluation on LTO of National project structural integrity assessment of reactor pressure vessels

Katsumata, Genshichiro; Masaki, Koichi*; Osakabe, Kazuya*; Nishikawa, Hiroyuki*; Katsuyama, Jinya; Nishiyama, Yutaka; Onizawa, Kunio

no journal, , 

To assure the structural integrity of a reactor pressure vessel (RPV) is known as one of the critical issues to maintain the safe long-term operation of a nuclear power plant. In Japan, the assessment methods for RPV integrity, stipulated in the codes and standards, have been endorsed by the regulatory body. Authors have initiated extensive research on the improvement of structural integrity assessment methods of RPVs. In this paper, we describe some research results obtained from the second-year activity. These include the study on revisiting the technical background of the methods, such as loading conditions, postulated crack definition, the other evaluation methods. In addition, studies on probabilistic methods for the applicability to the current rules and the standardization of the probabilistic analysis methods have been presented.

Oral presentation

Overview of researches at Structural Integrity Research Group

Katsuyama, Jinya; Nishiyama, Yutaka; Udagawa, Makoto; Yamaguchi, Yoshihito; Katsumata, Genshichiro

no journal, , 

no abstracts in English

Oral presentation

Effects of transient type and flaw density on through-wall cracking frequency of reactor pressure vessel under pressurized thermal shock events

Masaki, Koichi*; Osakabe, Kazuya*; Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio

no journal, , 

To assure the structural integrity of a reactor pressure vessel (RPV) is one of the most critical issues to maintain the safe long-term operation of a nuclear power plant. In Japan, the assessment methods for RPV integrity using deterministic fracture mechanics are provided in Japan Electric Association Code (JEAC). Meanwhile, a regulation on the fracture toughness requirements against PTS events based on a probabilistic fracture mechanics (PFM) analysis has been established in the U.S. In this paper, in order to apply probabilistic approach to domestic regulation, sensitivity analyses for flaw density or transient by reference to the data in the U.S. were performed using a PFM analysis code PASCAL3. We evaluated the effect of the flaw density or transient on through-wall cracking frequency (TWCF) and showed the specific example as a practical use of PFM.

Oral presentation

Study on the temperature condition in the pressurization thermal shock events based on three-dimensional thermal-hydraulics analysis

Katsumata, Genshichiro; Katsuyama, Jinya; Onizawa, Kunio; Nishiyama, Yutaka; Li, Y.

no journal, , 

no abstracts in English

Oral presentation

Study on structural integrity assessment of reactor pressure vessels

Uno, Shumpei; Katsuyama, Jinya; Katsumata, Genshichiro*; Masaki, Koichi*; Osakabe, Kazuya*; Li, Y.

no journal, , 

Assuring the structural integrity of a reactor pressure vessel (RPV) is known as one of the critical issues to maintain the safe long-term operation of a nuclear power plant. Authors have been conducting extensive research on the improvement of structural integrity assessment methods of RPVs. In this paper, we describe some research results obtained from the recent activity including the review on the technical background of the methods and study on probabilistic methods for the applicability to the current code and standard.

Oral presentation

Failure probability evaluation considering epistemic uncertainty on probabilistic fracture mechanics analysis

Osakabe, Kazuya*; Masaki, Koichi; Miyamoto, Yuhei*; Katsumata, Genshichiro*; Katsuyama, Jinya

no journal, , 

Probabilistic fracture mechanics (PFM) analysis is a useful methodology for quantitative evaluation because failure probabilities can be calculated considering uncertainties of material properties. The recent studies about the classification of uncertainties of inputs for PFM analyses are introduced. In this paper, we describe overseas study on the handling of epistemic uncertainty and aleatory uncertainty in PFM analysis considering reliability.

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