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Mori, Takashi*; Shimada, Takahiro*; Motegi, Haruki*; Kai, Satoru*; Otani, Akihito*; Yamamoto, Tomohiko; Yan, X.
Proceedings of the ASME 2025 Pressure Vessels & Piping Conference (PVP2025) (Internet), 9 Pages, 2025/07
Nakamura, Izumi*; Takito, Kiyotaka; Shimazu, Ryuya*; Okuda, Yukihiko; Sakai, Michiya*; Otani, Akihito*; Watakabe, Tomoyoshi; Okuda, Takahiro; Shibutani, Tadahiro*; Shiratori, Masaki*
Proceedings of the ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 9 Pages, 2024/07
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.
Mori, Takashi*; Shimada, Takahiro*; Kai, Satoru*; Otani, Akihito*; Yamamoto, Tomohiko; Yan, X.
Proceedings of the ASME 2023 Pressure Vessels and Piping Conference (PVP 2023) (Internet), 8 Pages, 2023/07
Mori, Takashi*; Shimada, Takahiro*; Kai, Satoru*; Otani, Akihito*; Yamamoto, Tomohiko; Yan, X.
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 8 Pages, 2023/05
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.
Morishita, Masaki; Otani, Akihito*; Nakamura, Izumi*; Watakabe, Tomoyoshi; Shibutani, Tadahiro*; Shiratori, Masaki*
Journal of Pressure Vessel Technology, 142(2), p.021203_1 - 021203_14, 2020/04
Times Cited Count：17 Percentile：61.35(Engineering, Mechanical)A Code Case in the framework of the Nuclear Codes and Standards of Japan Society of Mechanical Engineers (JSME) has been published to incorporate seismic design evaluation methodologies for piping systems by detailed inelastic response analysis and strain-based fatigue criteria as an alternative design rule to the current rule, in order to provide a more rational seismic design evaluation by taking directly the response reduction due to plasticity energy absorption into account. The Code Case provides two strain-based criteria; one is a limit to maximum amplitude of equivalent strain amplitude derived from detailed analysis and the other is a limit to the fatigue usage factor also based on the equivalent strain amplitude. The Code Case also provides a guideline for recommended standard analysis method. Discussions are provided on the safety margin and reliability of fatigue evaluation by the detailed inelastic response analysis provided in the Code Case.
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
Morishita, Masaki; Otani, Akihito*; Watakabe, Tomoyoshi; Nakamura, Izumi*; Shibutani, Tadahiro*; Shiratori, Masaki*
Proceedings of the ASME 2017 Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 7 Pages, 2017/07
Nakamura, Izumi*; Otani, Akihito*; Morishita, Masaki; Shiratori, Masaki*; Watakabe, Tomoyoshi; Shibutani, Tadahiro*
Proceedings of the ASME 2017 Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 10 Pages, 2017/07
Watakabe, Tomoyoshi; Nakamura, Izumi*; Otani, Akihito*; Morishita, Masaki; Shibutani, Tadahiro*; Shiratori, Masaki*
Proceedings of the ASME 2017 Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 10 Pages, 2017/07
To provide a more rational seismic design, a new Code Case for seismic design of piping is now under development in the framework of JSME Nuclear Codes and Standards. The Code Case incorporates a dynamic elastic-plastic analysis procedure by employing finite element analysis as an alternative to the current design analysis method of elastic assumption. To confirm the applicability of inelastic response analysis, benchmark analyses have been conducted.
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 the ASME 2013 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
Yamamoto, Tomohiko; Imai, Yoshiyuki; Marufuji, Takato; Akasaka, Naoaki; Yan, X.; Otani, Akihito*; Kai, Satoru*; Mori, Takashi*; Uoji, Tomoki*; Ueno, Tomohiro*; et al.
no journal, ,
no abstracts in English
