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.

Improve long periodic ground motion evaluation with the epicenter on the border between Fukushima and Ibaraki prefectures in seismic design of nuclear facilities

Kirita, Fumio; Tominaga, Masahiro; Yamazaki, Toshihiko; Seshimo, Kazuyoshi; Uryu, Mitsuru

JAEA-Research 2022-006, 61 Pages, 2023/02

JAEA-Research-2022-006.pdf:6.24MB

Nuclear Fuel Cycle Engineering Laboratories (NCL) has been observing ground motion for a long time. On the border from northern Ibaraki prefecture to Fukushima prefecture, inland crustal earthquakes occur less frequently until the 2011 off the Pacific coast of Tohoku Earthquake (hereinafter referred to as Tohoku Earthquake). After Tohoku Earthquake, aftershocks have become more frequent in this area, and in the Hamadori region of Fukushima earthquake that a remarkable long periodic component was observed in the NCL seismic observation record. Until now there were no such things that long periodic components were observed at the observation points near the epicenter of April 2011 Fukushima earthquake, but it was thought basin structure in deep basement around the NCL affected the propagation process to NCL by reflection survey result. As basement structure of NCL affected the seismic wave propagation process, the seismic wave repeatedly reflects and refracts. For that reason, long periodic components of seismic waves may be possibly amplified. In this study, in order to refine the long periodic ground motion evaluation, using a three dimensional ground structure model (3D model) that can reflect the shape of the deep basement structure around the NCL. When modeling 3D ground structure which has a width of about 80km and a length of about 110km and ranges from the epicenter area of April 2011 Fukushima earthquake to the northern coastal area of Ibaraki prefecture modeled, improved the optimum ground structure model using multiple observation records and performed simulation analysis.

Vibration test and fatigue test for failure probability evaluation method with integrated energy

Kinoshita, Takahiro*; Okamura, Shigeki*; Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi; Futagami, Satoshi; Fukasawa, Tsuyoshi*

Transactions of the 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 7 Pages, 2022/07

The seismic evaluation of key components such as reactor vessel is important for the Seismic Probabilistic Risk Assessment (S-PRA) in a Sodium-Cooled Fast Reactor (SFR). Many components were damaged by cumulative damage like fatigue damage during seismic ground motion. However, general evaluation method for key components under seismic ground motion has been based on static loads and elastic region of materials. More accurate evaluation method for S-PRA, which can evaluate the failure of key components such as reactor vessels, has been actually required. In this study, failure probability evaluation method with integrated energy was developed by comparing the energy with vibration tests and fatigue tests. Vibration tests were performed to evaluate integrated vibration energy at failure by energy balance equation and fatigue tests were performed to evaluate integrated vibration energy at failure based on experimental results of fatigue tests.

Places reached through 11-year academic society collaboration activities

Takada, Tsuyoshi

Nihon Jishin Kogakkai-Shi, (44), p.6 - 11, 2021/10

This report describes the status of nuclear seismic safety assurance in Japan before 2007, followed by the background and main achievements of the activities of the three research committees in collaboration with the Japan Association for Earthquake Engineering (JAEE) and the Atomic Energy Society of Japan (AESJ), and finally a summary of the points that the author considers important as the end of these activities.

Fundamental study on seismic safety margin for seismic isolated structure using the laminated rubber bearings

Fukasawa, Tsuyoshi*; Miyagawa, Takayuki*; Uchita, Masato*; Yamamoto, Tomohiko; Miyazaki, Masashi; Okamura, Shigeki*; Fujita, Satoshi*

Nihon Kikai Gakkai Rombunshu (Internet), 87(898), p.21-00007_1 - 21-00007_17, 2021/06

https://doi.org/10.1299/transjsme.21-00007

This paper describes a fundamental study on the seismic safety margin for the isolated structure using laminated rubber bearings. The variation of the seismic response assumed in the isolated structure will occur under the superposition of "Variations in seismic response due to input ground motions" and "Error with design value accompanying manufacture of the isolation devices ". The seismic response analysis which allows to their conditions is important to assess the seismic safety margin for the isolated structure. This paper clarifies that the seismic safety margin of the isolated structure, which consists of rubber bearings, for Sodium-cooled Fast Reactor (SFR) is ensured against the basis ground motions of Japan Electric Association Guide 4601 (JEAG4601) and SFR through the seismic response analysis considering the variation factors of seismic response. In addition, a relationship between the seismic safety margin and the excess probability of linearity limits is discussed using the results of seismic response analysis.

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.

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

https://doi.org/10.1115/1.4045133

Development of seismic counter measures against cliff edges for enhancement of comprehensive safety of nuclear power plants, 10; Avoidance of cliff edge for reactor vessel

Yamano, Hidemasa; Nishida, Akemi; Choi, B.; Takada, Tsuyoshi*

Transactions of the 25th International Conference on Structural Mechanics in Reactor Technology (SMiRT-25) (USB Flash Drive), 10 Pages, 2019/08

The objective of this study is to assess cliff edge effects, which are greatly important for nuclear power plants. Through assessments of failure probabilities (fragility), this study examined seismic margins of simulated two kinds of thin- and thick-walled reactor vessels by using response waveforms of the reactor building with/without a seismic isolation system obtained by seismic response analyses. The fragility analyses showed that the seismic isolation technology largely reduced the structural response effects nearly twice as much as that of the non-isolated plant. In focusing on uncertainty of response factor of components, the seismic isolation plant has a significant margin compared to the non-isolated plant even if factors from 0.5 to 2.0 are taken into account. This study concluded that the seismic isolation technology is effective to avoid cliff-edge effects.

Crack growth prediction for cracked dissimilar metal weld joint in pipe under large seismic cyclic loading

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

Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 8 Pages, 2018/07

In order to assess the structure integrity of cracked pipes considering occurrence of large earthquakes, crack growth evaluation method for cracked pipes is important. In present study, evaluation method of crack growth by seismic loading was proposed for a dissimilar metal weld joint of nickel based alloy through experimental study using small specimens. Then, validation of the proposed method was performed through crack growth tests by using dissimilar metal weld pipe with circumferential through-wall crack. The predicted crack growth values were in good agreement with the experimental results and the applicability of the proposed method was confirmed.

Uncertainty assessment of structural modeling in the seismic response analysis of nuclear facilities

Choi, B.; Nishida, Akemi; Muramatsu, Ken*; Takada, Tsuyoshi*

Transactions of the 24th International Conference on Structural Mechanics in Reactor Technology (SMiRT-24) (USB Flash Drive), 10 Pages, 2017/08

In order to clarify the influence of the modeling method on the result of seismic response analysis of nuclear facility, seismic response analysis using various simulated input ground motions was carried out and the uncertainty of response results were statistically analyzed. In particular, we focused on the difference of the response due to the structural modeling method (a conventional sway-rocking model and 3D FE model), and the relations among the input level, floor position, and response results were described and discussed.