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Journal Articles

Application of probabilistic fracture mechanics to reactor pressure vessel using PASCAL4 code

Lu, K.; Katsuyama, Jinya; Li, Y.; Yoshimura, Shinobu*

Journal of Pressure Vessel Technology, 143(2), p.021505_1 - 021505_8, 2021/04

 Times Cited Count:0

JAEA Reports

Guideline on seismic fragility evaluation for aged piping (Contract research)

Yamaguchi, Yoshihito; Katsuyama, Jinya; Masaki, Koichi*; Li, Y.

JAEA-Research 2020-017, 80 Pages, 2021/02

JAEA-Research-2020-017.pdf:3.5MB

The seismic probabilistic risk assessment (seismic PRA) is an important methodology to evaluate the seismic safety of nuclear power plants. Regarding seismic fragility evaluations performed in the seismic PRA, the Probabilistic Fracture Mechanics (PFM) can be applied as a useful evaluation technique for aged piping with crack or wall thinning due to the age-related degradation. Here, to advance seismic PRA methodology for the long-term operated nuclear power plants, a guideline for the fragility evaluation on the typical aged piping of nuclear power plants has been developed taking the aged-related degradation into account.

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

Stress intensity factor solutions for surface cracks with large aspect ratios in cylinders and plates

Zhang, T.; Lu, K.; Katsuyama, Jinya; Li, Y.

International Journal of Pressure Vessels and Piping, 189, p.104262_1 - 104262_12, 2021/02

 Times Cited Count:0 Percentile:100(Engineering, Multidisciplinary)

Journal Articles

Plasticity correction on stress intensity factor evaluation for underclad cracks in reactor pressure vessels

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

Journal of Pressure Vessel Technology, 142(5), p.051501_1 - 051501_10, 2020/10

 Times Cited Count:0 Percentile:100(Engineering, Mechanical)

Journal Articles

Constraint effect on fracture mechanics evaluation for an under-clad crack in a reactor pressure vessel steel

Shimodaira, Masaki; Tobita, Toru; Takamizawa, Hisashi; Katsuyama, Jinya; Hanawa, Satoshi

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

In JEAC 4206 which prescribes the methodology for assessing the structural integrity of reactor pressure vessels (RPVs), an under-clad crack (UCC) at the inner surface of RPV is postulated, and it is required that the fracture toughness of RPV steels is higher than stress intensity factor for at the crack tip during the pressurized thermal shock event. In the present study, to investigate the effect of cladding on the fracture toughness, we performed three-point bending fracture toughness tests and finite element analyses (FEAs) for an RPV steel containing an UCC or a surface crack, and the constraint effect for UCC was also discussed. As the result, we found that the fracture toughness for UCC was considerably higher than that for surface crack. On the other hand, the FEAs showed that the cladding decreased the constraint effect for UCC.

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

Extension of PASCAL4 code for probabilistic fracture mechanics analysis of reactor pressure vessel in boiling water reactor

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

Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 10 Pages, 2020/08

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

Recent verification activities on probabilistic fracture mechanics analysis code PASCAL4 for reactor pressure vessel

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

Mechanical Engineering Journal (Internet), 7(3), p.19-00573_1 - 19-00573_14, 2020/06

Journal Articles

Expansion of high temperature creep test data for failure evaluation of BWR lower head in severe accident

Yamaguchi, Yoshihito; Katsuyama, Jinya; Kaji, Yoshiyuki; Osaka, Masahiko; Li, Y.

Mechanical Engineering Journal (Internet), 7(3), p.19-00560_1 - 19-00560_12, 2020/06

Since the Fukushima Daiichi nuclear power plant accident, we have been developing a failure evaluation method that considers creep damage mechanisms using detailed three-dimensional finite element analysis model of lower head including penetration, stub tubes, and weld parts, etc., for the early completion of the decommissioning of the nuclear power plants in Fukushima Daiichi. For the finite element analysis, we have been obtaining material properties for which no data are provided in existing databases or in the literature. In particular, creep data corresponding to the high temperature region near the melting point of materials is important in evaluating creep deformation under severe accident conditions. In this study, we obtained the uniaxial tensile and creep properties for low-alloy steel, stainless steel, and Ni-based alloy. In particular, creep test data with long rupture times at high temperatures are expanded using a tensile test machine that can measure the elongation of test specimens in a noncontact measurement system. The parameters related to the failure evaluation were improved on the basis of the expanded creep database.

Journal Articles

Crack growth evaluation for cracked stainless and carbon steel pipes under large seismic cyclic loading

Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.; Onizawa, Kunio

Journal of Pressure Vessel Technology, 142(2), p.021906_1 - 021906_11, 2020/04

 Times Cited Count:1 Percentile:100(Engineering, Mechanical)

Journal Articles

Improvements on evaluation functions of a probabilistic fracture mechanics analysis code for reactor pressure vessels

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

Journal of Pressure Vessel Technology, 142(2), p.021208_1 - 021208_11, 2020/04

 Times Cited Count:2 Percentile:46.82(Engineering, Mechanical)

Journal Articles

Guideline on probabilistic fracture mechanics analysis for Japanese reactor pressure vessels

Katsuyama, Jinya; Osakabe, Kazuya*; Uno, Shumpei*; Li, Y.; Yoshimura, Shinobu*

Journal of Pressure Vessel Technology, 142(2), p.021205_1 - 021205_10, 2020/04

 Times Cited Count:0 Percentile:100(Engineering, Mechanical)

no abstracts in English

Journal Articles

Verification of safety margin of reactor pressure vessel exposed to various thermal transients based on probabilistic approach

Suzuki, Masahide*; Murakami, Kenta*; Suzuki, Takashi*; Okayama, Ryuta*; Katsuyama, Jinya; Li, Y.

E-Journal of Advanced Maintenance (Internet), 11(4), p.172 - 178, 2020/02

no abstracts in English

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:100(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

Ion-induced irradiation hardening of the weld heat-affected zone in low alloy steel

Ha, Yoosung; Takamizawa, Hisashi; Katsuyama, Jinya; Hanawa, Satoshi; Nishiyama, Yutaka

Nuclear Instruments and Methods in Physics Research B, 461, p.276 - 282, 2019/12

 Times Cited Count:0 Percentile:100(Instruments & Instrumentation)

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.

241 (Records 1-20 displayed on this page)