Katsuyama, Jinya; Osakabe, Kazuya*; Uno, Shumpei; Li, Y.; Yoshimura, Shinobu*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 9 Pages, 2017/07
A structural integrity assessment methodology based on probabilistic fracture mechanics (PFM) is a rational methodology in evaluating failure frequency of reactor pressure vessels (RPVs) by considering the probabilistic distributions of various influence factors related to the aged degradation. We have developed a PFM analysis code PASCAL to evaluate the failure frequency of RPVs considering the neutron irradiation embrittlement and pressurized thermal shock (PTS) events. We have also developed a guideline on the structural integrity assessment of RPVs based on PFM to improve the applicability of PFM in Japan and to be able to perform the PFM analyses and evaluate through-wall cracking frequency of RPVs. The technical basis for PFM analysis is provided and the latest knowledge is included in the guideline. In this paper, an overview of the guideline and some typical analysis results obtained based on the guideline and Japanese database related to PTS evaluation are presented.
Masaki, Koichi; Miyamoto, Yuhei*; Osakabe, Kazuya*; Uno, Shumpei*; Katsuyama, Jinya; Li, Y.
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 7 Pages, 2017/07
A probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency (JAEA). PASCAL can evaluate failure frequencies of Japanese reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events based on domestic structural integrity assessment models and data of influence factors. In order to improve the engineering applicability of PFM to Japanese RPVs, we have performed verification of the PASCAL. In general, PFM code consists of many functions such as fracture mechanics evaluation functions, probabilistic evaluation functions including random variables sampling modules and probabilistic evaluation models, and so on. The verification of PFM code is basically difficult because it is impossible to confirm such functions through the comparison with experiments. When a PFM code is applied for evaluating failure frequencies of RPVs, verification methodology of the code should be clarified and it is important that verification results including the region and process of the verification of the code are indicated. In this paper, our activities of verification for PASCAL are presented. We firstly represent the overview and methodology of verification of PFM code, and then, some verification examples are provided. Through the verification activities, the applicability of PASCAL in structural integrity assessments for Japanese RPVs was confirmed with great confidence.
Katsuyama, Jinya; Osakabe, Kazuya*; Uno, Shumpei; Li, Y.
JAEA-Research 2016-022, 40 Pages, 2017/02
For reactor pressure vessels (RPVs) in the light water reactors, the fracture toughness decreases due to the neutron irradiation embrittlement with operating years. In Japan, to prevent RPVs from a nil-ductile fracture, deterministic fracture mechanics methods in accordance with the codes provided by the Japan Electric Association are performed for assessing the structural integrity of RPVs under the pressurized thermal shock (PTS) events by taking the neutron irradiation embrittlement into account. On the other hand, in recent years, probabilistic methodologies for PTS evaluation are introduced into regulations in Europe and the United States. For example, in the United States, a PTS screening criterion related to the reference temperature derived by the probabilistic method is stipulated. If the screening criterion is not satisfied, it is approved to perform the evaluation based on the probabilistic method by calculating numerical index such as through-wall crack frequency (TWCF). To reach the objectives that persons who have knowledge on the fracture mechanics can carry out the PFM analyses and obtain TWCF for a domestic RPVs by referring to this report, we develop the guideline on a structural integrity assessment method based on PFM by reflecting the latest knowledge and expertise.
Okada, Hiroshi*; Koya, Hirohito*; Kawai, Hiroshi*; Li, Y.; Osakabe, Kazuya*
Engineering Fracture Mechanics, 158, p.144 - 166, 2016/06
The stress intensity factor (SIF) solutions of semi-elliptical cracks with high aspect ratios in plate and thick wall cylinder have been investigated under various assumed stress distributions. The authors have developed an automated analysis procedure to perform parametric studies on crack shapes and loading conditions. It consists of programs to perform automatic mesh generation, analysis execution including assignments of boundary conditions and SIF evaluations by virtual crack closure integral method. It was also found that SIF solutions for the thick wall cylinder and for the complex structure could be estimated by those for the flat plate.
Li, Y.; Hasegawa, Kunio; Katsumata, Genshichiro; Osakabe, Kazuya*; Okada, Hiroshi*
Journal of Pressure Vessel Technology, 137(5), p.051207_1 - 051207_8, 2015/10
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.
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.
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.
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 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.
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.
Katsuyama, Jinya; Ito, Hiroto*; Li, Y.*; Osakabe, Kazuya*; Onizawa, Kunio; Yoshimura, Shinobu*
International Journal of Pressure Vessels and Piping, 117-118, p.56 - 63, 2014/05
Several probabilistic fracture mechanical (PFM) analysis codes have been improved or developed in Japan, such as PASCAL-SP developed at JAEA, and PRAISE-JNES developed at JNES for structural integrity assessment of aged piping in nuclear power plants. Although they were developed for different purposes, they have similar functions. In this paper, in order to confirm the reliability and applicability of two PFM analysis codes, PASCAL-SP and PRAISE-JNES, benchmark analyses on piping failure probability have been carried out considering typical aging mechanisms, such as fatigue crack growth for piping materials in BWR plants Moreover, a criterion is proposed to judge whether the differences between the analysis results from two codes can be acceptable. Based on the proposed criterion, it is concluded that the analysis results of these two codes are in good agreements.
Kanto, Yasuhiro*; Onizawa, Kunio; Osakabe, Kazuya*; Yoshimura, Shinobu*
Proceedings of 10th International Workshop on the Integrity of Nuclear Components (ASINCO-10), p.189 - 194, 2014/04
The international Round Robin (RR) activity was performed in PFM sub-committees in JWES in conjunction with Korea and Taiwan research groups. The purposes of this program were to establish reliable procedures to evaluate fracture probability of reactor pressure vessels during pressurized thermal shock and to maintain the continuous cooperation among Asian institutes in the probabilistic approach to nuclear safety. The results of this work were summarized at the previous ASINCO workshop in 2010. Here, as the phase 2 of the program, a new international round robin activity is planned. This paper describes the outline of the problem. The aims of the program and the matters to be noticed is presented.
Li, Y.*; Ito, Hiroto*; Osakabe, Kazuya*; Onizawa, Kunio; Yoshimura, Shinobu*
International Journal of Pressure Vessels and Piping, 99-100, p.61 - 68, 2012/11
A benchmark analysis was conducted using two probabilistic fracture mechanics analysis codes for aged piping in nuclear power plants, in order to confirm their reliability and applicability. These analysis codes have been improved or developed in Japan for the structural integrity evaluation and risk assessment considering the age related degradation mechanisms. In the benchmark analysis, the primary loop recirculation system piping in the boiling water reactor was selected as the typical piping system and stress corrosion cracking and fatigue were taken into account as the typical aging mechanisms. Moreover, a criterion was proposed for judging whether the differences between analysis results from the two codes are acceptable. Based on the benchmark analysis results and numerical investigation, it was concluded that the analysis results of these two codes agree very well.
Ito, Hiroto*; Li, Y.*; Osakabe, Kazuya*; Onizawa, Kunio; Yoshimura, Shinobu*
Journal of Mechanical Science and Technology, 26(7), p.2055 - 2058, 2012/07
Probabilistic fracture mechanics is a rational methodology in the structural integrity evaluation and risk assessment for aged piping in nuclear power plants. Several probabilistic fracture mechanical analysis codes have been improved or developed in Japan. In this paper, to verify the reliability and applicability of two of these codes, a benchmark analysis was conducted using their basic functions in consideration of representative piping systems in nuclear power plants and typical aging mechanisms. Based on the analysis results, we concluded that the analysis results of these two codes are in good agreement.
Onizawa, Kunio; Masaki, Koichi; Osakabe, Kazuya*; Nishikawa, Hiroyuki*; Katsuyama, Jinya; Nishiyama, Yutaka
Nippon Hozen Gakkai Dai-9-Kai Gakujutsu Koenkai Yoshishu, p.374 - 379, 2012/07
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 first-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.
Nishikawa, Hiroyuki*; Osakabe, Kazuya*; Goto, Nobuhisa*; Suzuki, Hirokazu*; Katsuyama, Jinya; Onizawa, Kunio
Mizuho Joho Soken Giho (Internet), 4(1), 5 Pages, 2012/03
Stainless steel with 5 mm thickness is weld-overlay cladded on the inner surface reactor pressure vessels for protecting the vessel walls against the corrosion. Residual stress which corresponds to yield stress is generated near the cladding layer due to the weld-overlay cladding. However, there is no specific provision in the codes related to the cladding. Therefore, in order to confirm how the residual stress affects on the structural integrity of RPVs during pressurize thermal shock events, we evaluate the residual stress caused by the weld-overlay cladding and effects of the stress on the structural integrity. In this paper, the analysis method of welding residual stress of RPVs using finite element method and that of fracture mechanics considering postulated crack are introduced.
Nakajima, Kenji; Kawamura, Seiko; Kikuchi, Tatsuya; Nakamura, Mitsutaka; Kajimoto, Ryoichi; Inamura, Yasuhiro; Takahashi, Nobuaki; Aizawa, Kazuya; Suzuya, Kentaro; Shibata, Kaoru; et al.
Journal of the Physical Society of Japan, 80(Suppl.B), p.SB028_1 - SB028_6, 2011/05
AMATERAS is a cold-neutron disk-chopper spectrometer in MLF, J-PARC. The construction of main part of the spectrometer has been completed in spring of 2009. Soon after that, we have started the commissioning work on AMATERAS. The performance of AMATERAS has been examined by test experiments in the course of commissioning. In parallel to these works, we have started the user program on AMATERAS from December 2009 and we are getting scientific results from our spectrometer. In this presentation, we will report the current status of AMATERAS including the results of performance tests and some of examples of scientific outputs.
Masaki, Koichi; Nishikawa, Hiroyuki*; Osakabe, Kazuya*; Onizawa, Kunio
JAEA-Data/Code 2010-033, 350 Pages, 2011/03
The probabilistic fracture mechanics (PFM) analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed in JAEA. The PASCAL code evaluates the conditional probabilities of crack initiation and fracture of a reactor pressure vessel (RPV) under transient conditions such as pressurized thermal shock (PTS). Previous version of PASCAL (PASCAL ver.2) that was released in 2007 has many functions including the evaluation method for an embedded crack and conditional probabilities of crack initiation and fracture of a RPV, PTS transient database, inspection crack detection probability model and others. A generalized analysis method is available on the basis of the development of PASCAL ver.3 and sensitivity analysis results. Graphical user interface (GUI) including a generalized method and some functions of PFM have been also updated for PASCAL3. This report provides the user's manual, examples of analysis and theoretical background of PASCAL ver.3.
Ito, Hiroto; Kato, Daisuke*; Osakabe, Kazuya*; Nishikawa, Hiroyuki; Onizawa, Kunio
JAEA-Data/Code 2009-025, 135 Pages, 2010/03
As a part of the aging and structural integrity research for LWR components, new PFM (Probabilistic Fracture Mechanics) analysis code PASCAL-SP (PFM Analysis of Structural Components in Aging LWR - Stress Corrosion Cracking at Welded Joints of Piping) has been developed. This code evaluates the failure probabilities at welding lines of aged piping by a Monte Carlo method. PASCAL-SP treats stress corrosion cracking (SCC) in piping, including approaches of NISA and JSME FFS Code. The development of the code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the latest knowledge in the SCC assessment and fracture criteria of piping. In addition, the accuracy of flaw detection and sizing at in-service inspection and residual stress distribution were modeled based on experimental data and introduced into PASCAL-SP. This report provides the user's manual and theoretical background of the code.
Onizawa, Kunio; Osakabe, Kazuya
Proceedings of 2007 ASME Pressure Vessels and Piping Division Conference/8th International Conference on Creep and Fatigue at Elevated Temperatures (PVP 2007/CREEP-8) (CD-ROM), 7 Pages, 2007/07
During a pressurized thermal shock (PTS) event, the overlay cladding on the inner surface of reactor pressure vessel (RPV) is subjected to high tensile stress compared to base metal because of the difference in thermal expansion coefficients between cladding and base metal. To calculate a stress intensity factor for a postulated crack considering the stress discontinuity with the plastic yielding of cladding, the scheme developed previously has been incorporated into the PASCAL code for the structural integrity analysis. Using the new scheme, conditional probabilities of crack initiation (PCI) were calculated for a typical RPV with a surface crack or under-clad crack under some PTS transients. The PCI values were quantitatively evaluated as a function of neutron fluence using the PASCAL code. It is concluded that the new scheme reduces significantly the PCI value for a surface crack as compared with the conventional methods based on elastic stress analysis.
Osakabe, Kazuya; Onizawa, Kunio; Shibata, Katsuyuki; Suzuki, Masahide
Nippon Genshiryoku Gakkai Wabun Rombunshi, 6(2), p.161 - 171, 2007/06
As a part of the materials aging degradation and structural integrity research for LWR components, the probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed in JAEA. This code evaluates the conditional probabilities of crack initiation and fracture of a reactor pressure vessel (RPV) under transient conditions such as pressurized thermal shock (PTS). A standardized analysis method is proposed on the basis of the development of PASCAL ver.2 and results of sensitivity analyses. Graphical user interface (GUI) including the standardized analysis method as default settings and values has been also developed for PASCAL ver.2. A case study showed that non-destructive examination with good performance had a more significant effect on the probability of failure than non-destructive examinations repeated with low performance.