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Ichihara, Yoshitaka*; Nakamura, Naohiro*; Moritani, Hiroshi*; Choi, B.; Nishida, Akemi
Frontiers in Built Environment (Internet), 7, p.676408_1 - 676408_14, 2021/06
The objective of this study is the improvement of response evaluations of structures, facilities and equipment in evaluation of three-dimensional seismic behavior of nuclear power plant facilities, by three-dimensional finite element method model, including separation and sliding between the soil and the basement walls. To achieve this, simulation analyses of Kashiwazaki Kariwa nuclear power plant unit 7 reactor building under the 2007 Niigataken-chuetsu-oki earthquake event were carried out. These simulation analyses consider soil-structure interaction using a three-dimensional finite element method model in which the soil and building are three-dimensionally modeled by the finite element method. It is found that basemat uplift is generated on east side of the basemat edge, and this has an important influence on the results. The importance is evidenced by the difference of local response in soil pressure characteristics beneath the edge of basemat, the soil pressure characteristics along the east side of basement wall and the maximum acceleration response at the west end of the embedded surface. Although, in this particular study, basemat uplift, separation and sliding have only a relatively small influence on the maximum acceleration response of embedded surface and the soil pressure characteristics along the basement walls and beneath the basemat, under strong earthquake motion, these influences can be significant, therefore appropriate evaluation of this effect should be considered.
Choi, B.; Nishida, Akemi; Li, Y.; Muramatsu, Ken*; Takada, Tsuyoshi*
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07
After the 2011 Fukushima accident, nuclear power plants are required to take countermeasures against accidents beyond design basis conditions. In seismic probabilistic risk assessment (SPRA), uncertainty can be classified as either aleatory uncertainty, which cannot be reduced, or epistemic uncertainty, which can be reduced with additional knowledge and/or information. To improve the reliability of SPRA, efforts should be made to identify and reduce the epistemic uncertainty caused by the lack of knowledge. In this study, we focused on the difference in seismic response by modeling methods, which is related epistemic uncertainty. We conducted a seismic response analysis with two kinds of modeling methods; a three-dimensional finite-element model and a conventional sway-rocking stick model, by using simulated various input ground motions, which is related to aleatory uncertainty. And then we quantified the seismic floor response results of the various input ground motions of each modeling methods. For the uncertainty quantification related to different modeling methods, we further perform a statistical analysis of the floor response results of the nuclear reactor building. Finally, we discussed how to utilize the results from these calculations for the quantification of uncertainty in fragility analysis for SPRA.
Nakakura, Kensuke*; Yagenji, Akira*; Shoji, Teruaki*; Araki, Masanori; Neyatani, Yuzuru; Tanaka, Eiichi*; Shima, Hiroaki*
JAERI-Tech 2005-058, 61 Pages, 2005/09
JAERI-Tech-2005-058.pdf:4.49MB
The design of International Thermonuclear Experimental Reactor (ITER) had been done during the period of Engineering Design Activities (EDA) and its final deign report was published on July 2001. After that the further design studies have been performed as activities in the CTA and ITA periods. As for Tokamak complex, Japan Atomic Energy Research Institute (JAERI) performed preliminary study for both static and dynamic analyses with a few types of analysis models based on the Technical Guidelines for Aseismic Design of Nuclear Power Plants etc during EDA. In comparison with reactor building for nuclear fission power plant, its outside is not so different from reactor building's one. However, its structural characteristics inside building are apparently different from reactor building's one, because the bio shield of Tokamak complex has a lot of large penetration of ports. Therefore, characteristic of dynamic analysis models for Tokamak complex needs to be evaluated in order to perform the detailed design of Tokamak complex in account for site-dependent conditions. This report describes results of characteristic evaluation with representative analytical models for Tokamak complex, which are modeled with Finite Element Method and Lumped mass, and it is confirmed that adopted evaluation method of the bio shield wall and floor's stiffnesses around the bio shield wall for Lumped mass model is adequate.
A.Gilanyi*; Akiba, Masato; Okumura, Yoshikazu; Sugimoto, Makoto; Takahashi, Yoshikazu; Tsuji, Hiroshi
JAERI-Research 99-014, 71 Pages, 1999/03
JAERI-Research-99-014.pdf:3.56MB
no abstracts in English
Shimizu, Katsusuke*; ; Koizumi, Koichi; ; Nishio, Satoshi; Sasaki, Takashi*; Tada, Eisuke
JAERI-M 92-135, 139 Pages, 1992/09
no abstracts in English
; Suzuki, Hideyuki;
IWGGCR-22, p.60 - 66, 1990/00
no abstracts in English
; Harayama, Yasuo; ;
JAERI-M 8468, 77 Pages, 1979/09
no abstracts in English
