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Tsukimori, Kazuyuki; Ando, Masanori; Yada, Hiroki
Transactions of the 23rd International Conference on Structural Mechanics in Reactor Technology (SMiRT-23) (USB Flash Drive), 10 Pages, 2015/08
Multi-ply bellows is widely used because its flexibility and pressure tightness. However, the analytical treatment of its behaviours is difficult compared with that of single-ply bellows, since the uncertainty of friction between plies exists. In this paper the strength evaluation equations of multi-ply bellows are examined and the friction effect between plies is investigated through simplified models and detailed FEM analyses of two-ply bellows for example including the effect of material plasticity and internal pressure. Mainly following results are obtained. (1) The behaviours of two-ply bellows without friction can be analysed approximately by using single-ply bellows model even in the elastic-plastic region. The effort of numerical analysis can be reduced if two-ply bellows analyses can be replaced by the analyses of single-ply bellows model. (2) The effect of friction is relatively small up to about 0.4 by friction coefficient. Therefore, the behaviours of two-ply bellows can be evaluated practically ignoring friction effect. (3) The strength of multi-ply bellows against pressure, i.e. buckling, can be evaluated conservatively by assuming no friction, even if considered the internal pressure with friction. Therefore, the evaluation method without friction is recommended for design.
Takaya, Shigeru; Kamishima, Yoshio*; Machida, Hideo*; Watanabe, Daigo*; Asayama, Tai
Transactions of the 23rd International Conference on Structural Mechanics in Reactor Technology (SMiRT-23) (USB Flash Drive), 10 Pages, 2015/08
In our previous study, a new process for determination of in-service inspection (ISI) requirements was proposed on the basis of the System Based Code concept. The proposed process consists of two complementary evaluations, one focusing on structural integrity and the other on plant safety. In this study, ISI requirements for a reactor guard vessel and a core support structure of the prototype sodium-cooled fast breeder reactor in Japan, Monju, were investigated according to the proposed process. The proposed process is expected to contribute to realize effective and rational ISI by properly taking into account plant-specific features.
Nakajima, Norihiro; Nishida, Akemi; Kawakami, Yoshiaki; Suzuki, Yoshio; Matsukawa, Keisuke*; Oshima, Masami*; Izuchi, Hisao*
Transactions of the 23rd International Conference on Structural Mechanics in Reactor Technology (SMiRT-23) (USB Flash Drive), 10 Pages, 2015/08
The digital shaking table is introduced to carry out numerical experiments for the so called STRUCTURE of a petroleum plant. In numerical experiments, STRUCTURE was precisely modelled as it is designed and meshed into fine finite elements. The components of STRUCTURE were meshed one by one, and the code of a finite element analysis for structure of assembly gathered every meshed components to run time domain response analysis. Four waves are applied to the analysis to determine its behaviour. Four waves are namely as El Centro, Taft, Hachinohe, and Geiyo. The results of experiments are discussed by comparing accumulating data in the past. It is concluded to reconfirm the methodology of gathering meshed components and a finite element analysis for structure of assembly with the STRUCTURE.
Nishida, Akemi; Choi, B.; Itoi, Tatsuya*; Takada, Tsuyoshi*; Furuya, Osamu*; Muramatsu, Ken*
Transactions of the 23rd International Conference on Structural Mechanics in Reactor Technology (SMiRT-23) (USB Flash Drive), 10 Pages, 2015/08
This study focused on uncertainty-assessment frameworks and the development of relevant software to improve the reliability of seismic probabilistic risk assessment (SPRA) for NPP and promote its further use. This research aimed at contributing to development of implementation guidelines on epistemic uncertainty. Some sensitivity analyses were performed using a three dimensional reactor-building model and a conventional evaluation model by using 3D vibration simulator for NPP of JAEA, and the results were provided to the experts for expert-opinion elicitation. The results of the sensitivity analyses were related to the uncertainty evaluation of the buildings and soil to evaluate the fragility of the equipment. In this paper, those results will be shown in comparison with a conventional evaluation model.
Igarashi, Sayaka*; Sakamoto, Shigehiro*; Uchiyama, Yasuo*; Yamamoto, Yu*; Nishida, Akemi; Muramatsu, Ken; Takada, Tsuyoshi*
Transactions of the 23rd International Conference on Structural Mechanics in Reactor Technology (SMiRT-23) (USB Flash Drive), 10 Pages, 2015/08
In the preceding study, the methodology to generate ground-motion time histories for advanced PRA of NPPs was proposed by Nishida et al.. They are consistent with seismic hazard at reference site, and incorporate uncertainties of seismic-source characteristics. The ground motions utilized in conventional PRA are generated to fit to specified spectra such as UHS, and they are often generated without considering the variation of spectra. Even if it is considered, their inter-period correlations are generally assumed to be 1.0. In this paper, the authors prepared some cases of ground-motions sets. Ground motions are generated to fit to the response spectra calculated from hazard-consistent ground motions. While the target response spectra have the same median for all case, they have different variation and inter-period correlation. The response analyses of general RC structure and PWR building are conducted and the damage frequencies of simplified equipment system are compared.
Muta, Hitoshi*; Muramatsu, Ken*; Furuya, Osamu*; Uchiyama, Tomoaki*; Nishida, Akemi; Takada, Tsuyoshi*
Transactions of the 23rd International Conference on Structural Mechanics in Reactor Technology (SMiRT-23) (USB Flash Drive), 10 Pages, 2015/08
Seismic PRA is an effective measure to consider the countermeasures and improvement plans to secure the further safety of nuclear power plants regarding to seismic risk for the earthquake exceeding the reference ground motion. However, the application of the seismic PRA has not been utilized sufficiently so far. One of the reasons is that there is not enough consensus among stakeholders regarding to the evaluation methods and consideration of uncertainty for decision-making. This study proposes the mathematic framework to treat the uncertainty properly related to the evaluation of Core Damage Frequency induced by earthquake, the method to evaluate the fragility utilizing expert knowledge, the probabilistic model to cope with the aleatory uncertainty as well as the development of analyzing code including these considerations for the improvement of the reliability of the method and enhancement of utilization of the products of Seismic PRA.
Takada, Tsuyoshi*; Itoi, Tatsuya*; Nishida, Akemi; Furuya, Osamu*; Muramatsu, Ken*
Transactions of the 23rd International Conference on Structural Mechanics in Reactor Technology (SMiRT-23) (USB Flash Drive), 10 Pages, 2015/08
The assessment of seismic safety of nuclear power plant facilities has been performed by identifying and quantifying uncertainties in seismic probabilistic risk assessment (SPRA). For the evaluation process, all uncertainties are classified into either aleatory or epistemic uncertainty for their quantification. In this study, systematic evaluation of the epistemic uncertainty on the seismic fragility of structures and equipment is studied and implemented for a model NPP. There are two expert groups formed in this project: experts in the field of buildings and soil ground (CE experts) and experts in the field of pipe and equipment (ME experts). Along with ample results from relevant sensitivity analyses conducted, elicited opinions are carefully treated and classified into several specific areas and integrated into the form of knowledge tree technique (KTT), all of which can be utilized for future fragility estimation.