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Sugino, Hideharu*; Ito, Hiroto*; Onizawa, Kunio; Suzuki, Masahide

Nihon Genshiryoku Gakkai Wabun Rombunshi, 4(4), p.233 - 241, 2005/12

The purpose of this research is to establish the reliability evaluation method of aged nuclear power components for seismic events from a viewpoint of long-term use of the existing light water reactor nuclear power plants. For this purpose, we developed a piping failure probability evaluation code "PASCAL-SC" based on probabilistic fracture mechanics, and a probabilistic seismic hazard evaluation code "SHEAT-FM" for calculating the seismic occurrence probability of a plant site, paying attention to aging such as fatigue crack progress by the stress corrosion cracking and seismic load in primary coolant piping system. We proposed the reliability evaluation method of aged piping for seismic events by combination of these codes. Using this method, we evaluated the reliability of a weld line in the PLR(Primary Loop Recirculation system) piping of the BWR model plant for seismic events.

Moriyama, Kiyofumi; Takagi, Seiji; Muramatsu, Ken; Nakamura, Hideo; Maruyama, Yu

Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 9 Pages, 2005/05

The containment failure probability due to ex-vessel steam explosions were evaluated for a BWR Mk-II model plant. The evaluation was made for two scenarios: a steam explosion in the pedestal area, or in the suppression pool. A probabilistic approach, Latin Hypercube Sampling (LHS), was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The fragility curves connecting the steam explosion loads and containment failure were developed based on simplified assumptions on the containment failure scenarios. The mean conditional probabilities of containment failure per occurrence of a steam explosion were for suppression pool and for pedestal area. Note that the results depend on the assumed range of input parameters and fragility curves that involve conservatism and simplification.

Ishikura, Shuichi*; Shiga, Akio*; Futakawa, Masatoshi; Kogawa, Hiroyuki; Sato, Hiroshi; Haga, Katsuhiro; Ikeda, Yujiro

JAERI-Tech 2005-026, 65 Pages, 2005/03

Failure probability analysis was carried out to estimate the lifetime of the mercury target which will be installed into the JSNS (Japan spallation neutron source) in J-PARC (Japan Proton Accelerator Research Complex). The lifetime was estimated as taking loading condition and materials degradation into account. Considered loads imposed on the target vessel were the static stresses due to thermal expansion and static pre-pressure on He-gas and mercury and the dynamic stresses due to the thermally shocked pressure waves generated repeatedly at 25 Hz. Materials used in target vessel will be degraded by the fatigue, neutron and proton irradiation, mercury immersion and pitting damages, etc. The imposed stresses were evaluated through static and dynamic structural analyses. The material-degradations were deduced based on published experimental data. As results, it was quantitatively confirmed that the failure probability for the lifetime expected in the design is very much lower, 10 in the safety hull, meaning that it will be hardly failed during the design lifetime. On the other hand, the beam window of mercury vessel suffered with high-pressure waves exhibits the failure probability of 12%. It was concluded, therefore, that the leaked mercury from the failed area at the beam window is adequately kept in the space between the safety hull and the mercury vessel to detect mercury-leakage sensors.

Shibata, Katsuyuki; Onizawa, Kunio; Li, Y.*; Kato, Daisuke*

International Journal of Pressure Vessels and Piping, 81(9), p.749 - 756, 2004/09

Times Cited Count：5 Percentile：35.81(Engineering, Multidisciplinary)The paper describes the procedure to evaluate the ductile crack extension, where an increase in fracture resistance by a ductile crack extension is considered. Two standard -resistance curves are prepared for applying the elasto-plastic fracture criterion. Case studies concerning the effect of elasto-plastic fracture criterion were carried out using a severe PTS transient. The introduction of the elasto-plastic fracture criterion significantly contributes to remove the over-conservatism in applying the linear elastic fracture criterion. It was also found that the algorithm of the re-evaluation of crack tip characterization also has a significant effect on the failure probability.

Shibata, Katsuyuki; Onizawa, Kunio; Li, Y.*; Kanto, Yasuhiro*; Yoshimura, Shinobu*

Pressure Vessel and Piping Codes and Standards (PVP-Vol.480), p.235 - 242, 2004/00

A screening standard of small flaws that have no significant influence on the structural integrity is prescribed in ASME Code Sec.XI. From the viewpoint of probabilistic methodology, there are some concerns that weather or not the failure probability is uniform for flaws with various aspect ratios and failure frequencies are small enough. Moreover, acceptable flaws may be determined more rationally based the failure probability. A study was performed on the failure probability of RPVs with a surface flaw specified in Sec.XI using the PFM code PASCAL. A PTS transient of NRC/EPRI Benchmark Study was used. Analysis results showed that the conditional failure probability of a RPV with an initial flaw of acceptable depth depended on the aspect ratio. For a circular flaw, the failure probability is high due to the increase in crack initiation at the surface point. The maximum difference in failure probability reaches one order of magnitude. A case study for determining the acceptable flaws based on failure probability was also carried out.

Shibata, Katsuyuki; Kanto, Yasuhiro*; Yoshimura, Shinobu*; Yagawa, Genki*

Proceedings of 5th International Workshop on the Integrity of Nuclear Components, p.99 - 117, 2004/00

no abstracts in English

Uchiyama, Tomoaki; Oikawa, Tetsukuni; Kondo, Masaaki; Watanabe, Yuichi*; Tamura, Kazuo*

JAERI-Data/Code 2002-011, 205 Pages, 2002/03

This report is a user's manual of seismic system reliability analysis code SECOM2 developed at the JAERI for system reliability analysis, which is one of the tasks of seismic probabilistic safety assessment (PSA) of nuclear power plants (NPPs). The SECOM2 code has many functions such as calculation of component and system failure probabilities for given seismic motion levels at the site of an NPP based on the response factor method, calculation of accident sequence frequencies and the core damage frequency (CDF), importance analysis using various indicators, uncertainty analysis, and calculation of the CDF taking into account the effect of the correlations of responses and capacities of components. These analyses require the fault tree (FT) representing the occurrence condition of the system failures and core damage, information about responses and capacities of the components which compose the FT, and seismic hazard curve for the NPP site as input. This report presents calculation method used in the SECOM2 code and how to use those functions in the SECOM2 code.

Shibata, Katsuyuki; Onizawa, Kunio; Li, Y.*; Kato, Daisuke*

Proceedings of 4th International Workshop on the Integrity of Nuclear Components, p.31 - 41, 2002/00

no abstracts in English

Shibata, Katsuyuki; Kato, Daisuke*; Li, Y.*

Nuclear Engineering and Design, 208(1), p.1 - 13, 2001/08

Times Cited Count：20 Percentile：80.57(Nuclear Science & Technology)no abstracts in English

Li, Y.*; Kato, Daisuke*; Shibata, Katsuyuki; Onizawa, Kunio

International Journal of Pressure Vessels and Piping, 78(4), p.271 - 282, 2001/04

Times Cited Count：5 Percentile：42.81(Engineering, Multidisciplinary)no abstracts in English

Shibata, Katsuyuki; Onizawa, Kunio; Li, Y.*; Kato, Daisuke*

JAERI-Data/Code 2001-011, 233 Pages, 2001/03

no abstracts in English

Li, Y.*; Kato, Daisuke*; Shibata, Katsuyuki

JSME International Journal, Series A, 44(1), p.130 - 137, 2001/01

no abstracts in English

Shibata, Katsuyuki; Kato, Daisuke*; Li, Y.*

Emerging Technologies: Risk Assessment, Computational Mechanics and Advanced Engineering Topics (PVP-Vol.400), p.49 - 54, 2000/00

no abstracts in English

Y.Li*; Kato, Daisuke*; Shibata, Katsuyuki

Proceedings of 7th International Conference on Nuclear Engineering (ICONE-7) (CD-ROM), 10 Pages, 1999/00

no abstracts in English

*; Matsumoto, Kiyoshi; Hoshina, Hirofumi*

PSA95: Proc. of Probabilistic Safety Assessment Methodology and Applications, 2, p.735 - 740, 1995/00

no abstracts in English