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Nakamura, Izumi*; Otani, Akihito*; Okuda, Yukihiko; Watakabe, Tomoyoshi; Takito, Kiyotaka; Okuda, Takahiro; Shimazu, Ryuya*; Sakai, Michiya*; Shibutani, Tadahiro*; Shiratori, Masaki*
Dai-10-Kai Kozobutsu No Anzensei, Shinraisei Ni Kansuru Kokunai Shimpojiumu (JCOSSAR2023) Koen Rombunshu (Internet), p.143 - 149, 2023/10
In 2019, the JSME Code Case for seismic design of nuclear power plant piping systems was published. The Code Case provides the strain-based fatigue criteria and detailed inelastic response analysis procedure as an alternative design rule to the current seismic design, which is based on the stress evaluation by elastic response analysis. In 2022, it was approved to revise the Code Case with improving the cycle counting method for fatigue evaluation to the Rain flow method. In addition, the discussion to incorporate the elastic-plastic behavior of support structures is now in progress for the next revision of the Code Case. This paper discusses the contents and background of the 2022 revision, the progress of the next revision, and future tasks.
Furuya, Osamu*; Fujita, Satoshi*; Muta, Hitoshi*; Otori, Yasuki*; Itoi, Tatsuya*; Okamura, Shigeki*; Minagawa, Keisuke*; Nakamura, Izumi*; Fujimoto, Shigeru*; Otani, Akihito*; et al.
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 6 Pages, 2021/07
Since the Fukushima accident, with the higher safety requirements of nuclear facilities in Japan, suppliers, manufacturers and academic societies have been actively considering the reconstruction of the safety of nuclear facilities from various perspectives. The Nuclear Regulation Authority has formulated new regulatory standards and is in operation. The new regulatory standards are based on defense in depth, and have significantly raised the levels of natural hazards and have requested to strengthen the countermeasures from the perspective of preventing the simultaneous loss of safety functions due to common factors. Facilities for dealing with specific serious accidents are required to have robustness to ensure functions against earthquakes that exceed the design standards to a certain extent. In addition, since the probabilistic risk assessment (PRA) and the safety margin evaluation are performed to include the range beyond the design assumption in the safety improvement evaluation, it is very important to extent the special knowledge in the strength of important equipment for seismic safety. This paper summarizes the research and examination results of specialized knowledge on the concept of maintaining the functions of important seismic facilities and the damage index to be considered by severe earthquakes. In the other paper, the study on reliability of seismic capacity analysis for important equipment in nuclear facilities will be reported.
Otani, Akihito*; Kai, Satoru*; Kaneko, Naoaki*; Watakabe, Tomoyoshi; Ando, Masanori; Tsukimori, Kazuyuki*
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 10 Pages, 2018/07
This paper demonstrates an application result of the JSME Seismic Code Case to an actual complex piping system. The secondary coolant piping system of Japanese Fast Breeder Reactor, Monju, was selected as a representative of the complex piping systems. The elastic-plastic time history analysis for the piping system was performed and the piping system has been evaluated according to the JSME Seismic Code Case. The evaluation by the Code Case provides a reasonable result in terms of the piping fatigue evaluation that governs seismic integrity of piping systems.
Hirota, Noriaki; Terada, Atsuhiko; Yan, X.; Tanaka, Kohei*; Otani, Akihito*
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 7 Pages, 2018/07
Watakabe, Tomoyoshi; Tsukimori, Kazuyuki; Otani, Akihito*; Moriizumi, Makoto; Kaneko, Naoaki*
Mechanical Engineering Journal (Internet), 3(3), p.16-00054_1 - 16-00054_11, 2016/06
It is important to investigate the failure mode and ultimate strength of piping components in order to evaluate the seismic integrity of piping. Many failure tests of thick wall and high pressure piping for Light Water Reactors (LWRs) have been conducted, and the results suggest that the failure mode that should be considered in the design of a thick wall piping for LWRs under seismic loading is low cycle fatigue. On the other hand, Sodium cooled Fast Reactors (SFRs) is thin wall when compared to LWRs piping. Failure tests of a thin wall piping are necessary because past failure tests for LWRs piping are not enough to discuss failure behavior of a thin wall piping. Therefore, this present work investigated the failure mode and the ultimate strength of thin wall tees.
Watakabe, Tomoyoshi; Tsukimori, Kazuyuki; Otani, Akihito*; Moriizumi, Makoto; Kaneko, Naoaki*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05
Watakabe, Tomoyoshi; Tsukimori, Kazuyuki; Otani, Akihito*; Moriizumi, Makoto; Kaneko, Naoaki*
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 8 Pages, 2014/07
In recent years, earthquakes over design condition were observed in Japan. Confirming the ultimate strength and design safety margin of mechanical components is important for the seismic integrity. This study focused on piping components, and it was one of the most important mechanical components for protecting boundary of coolant. Failure tests of thick-walled piping components for Light Water Reactors (LWRs) described previously in the literature. According to these tests, the failure mode of thick-walled piping components under seismic cyclic loading was low cycle fatigue. However, failure tests have scarcely been performed on thin-walled piping components pressurized at low levels for Fast Breeder Reactors (FBRs). This paper presents dynamic failure tests of thin-walled piping components in FBRs. Based on the test results, the failure mode, the ultimate strength, and the elastic-plastic behavior are discussed.
Watakabe, Tomoyoshi; Kaneko, Naoaki*; Aida, Shigekazu*; Otani, Akihito*; Tsukimori, Kazuyuki; Moriizumi, Makoto; Kitamura, Seiji
Dynamics and Design Conference 2013 (D&D 2013) Koen Rombunshu (USB Flash Drive), 8 Pages, 2013/08
The piping in a nuclear power plant is laid across multiple floors of a single building or two buildings, which are supported at many anchors. As the piping is excited by multiple inputs from the supporting anchors during an earthquake, seismic response analysis by multiple excitations is needed to obtain the exact seismic response of the piping. However, few tests involving such multiple excitations have been performed to verify the validity of multiple excitation analysis. To perform rational seismic design and evaluation, it is important to investigate the seismic response by multiple excitations and verify the validity of the analysis method by multiple excitation test. This paper reports on the result of the shaking test using triple uni-axial shaking tables and a 3-dimensional piping model.
Watakabe, Tomoyoshi; Kaneko, Naoaki*; Aida, Shigekazu*; Otani, Akihito*; Moriizumi, Makoto*; Tsukimori, Kazuyuki; Kitamura, Seiji
Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 8 Pages, 2013/07
The piping in a nuclear power plant is laid across multiple floors of a single building or two buildings, which are supported at many points. As the piping is excited by multiple inputs from the supporting points during an earthquake, seismic response analysis by multiple excitations is needed to obtain the exact seismic response of the piping. However, few experiments involving such multiple excitations have been performed to verify the validity of multiple excitation analysis. To perform rational seismic design and evaluation, it is important to investigate the seismic response by multiple excitations and verify the validity of the analysis method by multiple excitation test. This paper reports on the result of the shaking test using triple uni-axial shaking tables and a 3-dimensional piping model.
Shimada, Takahiro*; Otani, Akihito*; Iwamoto, Kosuke*; Kitamura, Seiji
Nihon Kikai Gakkai Rombunshu, C, 77(777), p.1661 - 1673, 2011/05
Three dimensional seismic isolation devices have been developed for the base isolation system of the Fast Breeder Reactor that is an advanced nuclear power buildings. The developed seismic isolation system consists of the hydraulic type vertical springs with rocking suppression mechanism and the laminated rubber bearings for horizontal direction. In this paper, it is reported the frictional characteristics on high hydraulic pressure condition from the experiments on the 1/2 size of real device and the results of the seismic simulation on the real size building with isolation-device that has those characteristics.
Kunitomi, Kazuhiko; Shinozaki, Masayuki; ; Okubo, Minoru; Baba, Osamu; *; Otani, Akihito*
JAERI-M 92-147, 77 Pages, 1992/10
no abstracts in English
Kaneko, Naoaki*; Otani, Akihito*; Waki, Minoru*; Ogawa, Fujio*; Tsukimori, Kazuyuki
no journal, ,
no abstracts in English
Kaneko, Naoaki*; Kitamura, Seiji; Jimbo, Noboru*; Mizutani, Takumi*; Otani, Akihito*
no journal, ,
no abstracts in English
Suzuki, Tetsu*; Maruyama, Shigeki*; Kamijo, Chiharu*; Otani, Akihito*; Kan, Norio*; Yan, X.; Sato, Hiroyuki; Tazawa, Yujiro; Ohashi, Hirofumi; Tachibana, Yukio
no journal, ,
no abstracts in English
Yamamoto, Tomohiko; Yan, X.; Shimada, Takahiro*; Motegi, Haruki*; Kai, Satoru*; Otani, Akihito*
no journal, ,
no abstracts in English
Morishita, Masaki; Otani, Akihito*
no journal, ,
The JSME Code Case N-CC-008 provides design rules for piping seismic design by inelastic response analysis and strain-based fatigue criteria. The Code Case uses the Rainflow method for cycle counting. The Rainflow method is used for identifying pairs of peaks/valleys of the representative strain and their occurrence times. On the other hand, ASME Code reads "The Rainflow cycle counting method is recommended but not applicable for non-proportional loading." The authors analyzed the behavior of multi-axial strain of piping systems to test the non-proportionality. The analysis results showed that the multi-axial strain in piping systems induced by seismic excitation is almost proportional. Based on this result, the authors concluded that the Rainflow cycle counting method is well applicable for the piping seismic fatigue evaluation. In addition to these numerical analyses with the IS method, some discussions are also made in this paper from the viewpoint of the relation between vibration mode and strain components to support the conclusion that the piping strain by seismic loads is proportional.
Nakamura, Izumi*; Otani, Akihito*; Morishita, Masaki; Okuda, Yukihiko; Watakabe, Tomoyoshi; Shibutani, Tadahiro*; Takito, Kiyotaka; Okuda, Takahiro; Shiratori, Masaki*
no journal, ,
no abstracts in English
Morishita, Masaki; Otani, Akihito*
no journal, ,
no abstracts in English
Takito, Kiyotaka; Nakamura, Izumi*; Okuda, Yukihiko; Sakai, Michiya*; Shimazu, Ryuya*; Otani, Akihito*; Watakabe, Tomoyoshi; Okuda, Takahiro; Shibutani, Tadahiro*; Shiratori, Masaki*
no journal, ,
The Piping systems in the nuclear power plants are known to exhibit significant elastic plastic behavior before failure caused by the response displacement under serve seismic load. Current Code Case considers the elastic-plastic behavior of piping system itself, but assumes the elastic behavior for the support structure. Thus, it is desirable to incorporate the inelastic behavior of piping support structure as the future upgrading of the Code Case. In order to develop an evaluation method for inelastic behavior of support structures, the authors as the JSME task group carried out benchmark analyses of piping support structures. Then, the purpose of the benchmark analysis is in order to evaluate the influence of the analysis parameters in the elastic plastic analysis of piping support structures and to obtain knowledge on the variability of the analysis results. This paper reports on the outline and progress of the benchmark analyses.
山本 智彦; Yan, X.
島田 貴弘*; 大谷 章仁*; 甲斐 聡流*
JP, 2020-095752 
Patent licensing information
【課題】地震に対する浮体式構造物および搭載機器の励振を低減する。 【解決手段】浮体式免震システム100は、液体114を貯留する液体貯留部110と、液体114に浮揚して配置される浮体式構造物120と、液体114と接触する箇所に設けられ、気体134を収容する気体収容空間132と、を備え、液体114と気体134を含む流体中を伝播する地震波により応答する系の固有振動数に基づいて、気体収容空間132の体積が設定されている。
