Development of failure mitigation technologies for improving resilience of nuclear structures, 6; Resilience improvements of fast reactors by failure mitigation for excessive earthquake

Yamano, Hidemasa; Futagami, Satoshi; Ando, Masanori; Kurisaka, Kenichi

Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 11 Pages, 2024/03

In this study, the dynamic structural analysis of the reactor vessel for excessive earthquake using the FINAS/STAR code has shown the elephant foot buckling deformation and calculated the cumulative fatigue failure fraction. Using the calculation results, this paper describes the fragility curve using the safety factor method, indicating the significantly improved curve compared the previous one.

Development of probabilistic analysis code for evaluating seismic fragility of aged pipes with wall-thinning

Yamaguchi, Yoshihito; Nishida, Akemi; Li, Y.

Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 7 Pages, 2022/07

https://doi.org/10.1115/PVP2022-84009

The wall-thinning is one of the most important age-related degradation phenomena in nuclear piping systems. Furthermore, in recent years, several nuclear power plants in Japan have experienced severe earthquakes. Therefore, failure probability analysis and fragility evaluation of piping systems, taking both wall-thinning and seismic response stresses into consideration, have become increasingly important in seismic probabilistic risk assessment. In Japan Atomic Energy Agency, in order to evaluate the failure probability of aged piping system with wall-thinning, a probabilistic analysis code PASCAL-EC was developed. In this study, to evaluate the seismic fragility of a wall-thinned pipe, a model of seismic response stress considering the wall-thinning effect, a failure evaluation method for wall-thinned pipes, and functions related to uncertainties treatment for important influence parameters have been introduced to PASCAL-EC. In this paper, the improved PASCAL-EC is outlined and preliminary results of the seismic fragility evaluation performed using this code are provided.

Development of guideline on seismic fragility evaluation for aged piping

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

Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 9 Pages, 2021/07

The seismic probabilistic risk assessment is an important methodology to evaluate the seismic safety of nuclear power plants. In this assessment, the core damage frequency is evaluated from the seismic hazard, seismic fragilities, and accident sequence. Regarding the seismic fragility evaluation, the probabilistic fracture mechanics can be applied as a useful evaluation technique for aged piping systems with crack or wall thinning due to the age-related degradation mechanisms. In this study, to advance seismic probabilistic risk assessment methodology of nuclear power plants that have been in operation for a long time, a guideline on the seismic fragility evaluation of the typical aged piping systems of nuclear power plants has been developed considering the age-related degradation mechanisms. This paper provides an outline of the guideline and several examples of seismic fragility evaluation based on the guideline and utilizing the probabilistic fracture mechanics analysis code.

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.

Seismic capacity evaluation of a group of vertical U-tube heat exchanger with support frames for seismic PSA

Watanabe, Yuichi*; Muramatsu, Ken; Oikawa, Tetsukuni

Nuclear Engineering and Design, 235(23), p.2495 - 2512, 2005/12

https://doi.org/10.1016/j.nucengdes.2005.05.037

This paper presents an evaluation of seismic capacity of a group of vertical U-tube type heat exchangers(HXs) with support frames for residual heat removal systems of BWRs for seismic Probabilistic Safety Assessment in Japan. The median capacity was evaluated by a time history response analysis with a detailed model for a representative HX selected from four HXs. The logarithmic standard deviation(LSD) for uncertainty due to lack of knowledge was evaluated with consideration of the variabilities in three influential parameters, i.e., diameter of anchor bolt, weight of HX and position of center of gravity of HX. The dominant failure mode of HXs was the failure of anchor bolts of lugs mainly due to shearing stress. The capacity expressed in terms of zero period acceleration on the foundation of HX was evaluated to be 4,180 Gal(4.3 g) for median, LSD for uncertainty due to randomness was 0.11 from the variability in material property and LSD due to lack of knowledge was 0.21 to 0.53 depending on combination of the variability in design parameters to be considered.

Seismic reliability evaluation of electrical power transmission systems and its effect on core damage frequency

Oikawa, Tetsukuni; Fukushima, Seiichiro*; Takase, Hidekazu*; Uchiyama, Tomoaki*; Muramatsu, Ken

Transactions of 16th International Conference on Structural Mechanics in Reactor Technology (SMiRT-16) (CD-ROM), 8 Pages, 2001/08

no abstracts in English

Fragility evaluation of wall-thinned pipes using PASCAL-EC

Ebine, Noriya; Yamaguchi, Yoshihito; Katsuyama, Jinya; Nishida, Akemi; Li, Y.

no journal, , 

A probabilistic analysis code PASCAL-EC (PFM Analysis of Structural Components in Aging LWR - Erosion and Corrosion) have been improved for evaluating structural integrity of wall thinned pipes on the basis of the Monte Carlo method. In order to perform the fragility evaluation for wall-thinned pipes, several functions are introduced to PASCAL-EC in this work, such as probabilistic evaluation model to consider the uncertainty of seismic response stress, a simple evaluation model to considered the increase of the seismic stress due to wall thinning and failure evaluation methods for wall-thinned pipes. This paper provides some details of these models and preliminary results of fragility evaluation using improved PASCAL-EC code for wall-thinned pipes in nuclear power plants.