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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.
Takeda, Nobukazu; Kakudate, Satoshi; Shibanuma, Kiyoshi
Purazuma, Kaku Yugo Gakkai-Shi, 81(4), p.312 - 316, 2005/04
The vibration experiments of the support structures with flexible plates for the ITER major components such as toroidal field coil and vacuum vessel were performed using small-sized flexible plates aiming to obtain its basic mechanical characteristics such as dependence of the stiffness on the loading angle. The experimental results obtained by the hammering and frequency sweep tests were agreed each other, so that the experimental method is found to be reliable. In addition, the experimental results were compared with the analytical ones in order to estimate an adequate analytical model. As a result, the bolt connection strongly affected on the stiffness of the support structure. After studies of modeling the bolts, it is found that the analytical results modeling the bolts with finite stiffness only in the axial direction and infinite stiffness in the other directions agree well with the experimental ones. Using this model, the stiffness of the support structure for the ITER major components can be calculated precisely in order to estimate the dynamic behaviors.
Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka
JAERI-Tech 2004-072, 43 Pages, 2005/01
JAERI-Tech-2004-072.pdf:6.06MB
The vibration experiments of the support structures with flexible plates for the ITER major components such as the vacuum vessel (VV) and the toroidal field (TF) coil were performed aiming to obtain its basic mechanical characteristics. Based on the experimental results, numerical analysis regarding the actual support structure was performed and a simplified model of the support structure was proposed. A support structure was modeled by only two spring elements. The stiffness calculated by the spring model agrees well with that of shell model, simulating actual structures based on the experimental results. It is therefore found that the spring model with the only two values of stiffness enables to simplify the complicated support structure with flexible plates. Using the spring model, the dynamic analysis of the VV and TF coil were performed to estimate the integrity under the design earthquake. As a result, the maximum relative displacement of 8.6 mm between VV and TF coil is much less than designed clearance, 100 mm, so that the integrity of the components is ensured.
Takeda, Nobukazu; Omori, Junji*; Nakahira, Masataka
JAERI-Tech 2004-068, 27 Pages, 2004/12
JAERI-Tech-2004-068.pdf:7.68MB
ITER vacuum vessel (VV) is a safety component confining radioactive materials such as tritium and activated dust. An independent VV support structure with multiple flexible plates located at the bottom of VV lower port is proposed. This independent concept has two advantages: (1) thermal load due to the temperature deference between VV and the lower temperature components such as TF coil becomes lower and (2) the other components such as TF coil is categorized as a non-safety component because of its independence from VV. Stress analyses have been performed to assess the integrity of the VV support structure using a precisely modeled VV structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coil is found to be 15 mm, much less than the current design value of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME. Based on these assessments, the feasibility of the proposed independent VV support has been verified as a VV support.
Takeda, Nobukazu; Omori, Junji*; Nakahira, Masataka; Shibanuma, Kiyoshi
Journal of Nuclear Science and Technology, 41(12), p.1280 - 1286, 2004/12
Times Cited Count：3 Percentile：23.52(Nuclear Science & Technology)ITER vacuum vessel (VV) is a safety component confining radioactive materials. An independent VV support structure located at the bottom of VV lower port is proposed as an alternative concept, which is deferent from the current reference, i.e., the VV support is directly connected to the toroidal coil (TF coil). This independent concept has two advantages comparing to the reference one: (1) thermal load becomes lower and (2) the TF coil is categorized as a non-safety component. Stress Analyses have been performed to assess the integrity of the VV support structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coil is found to be 15 mm, much less than the current design value of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME, respectively. Based on these assessments, the feasibility of the proposed independent VV support has been verified as an alternative VV support.
Takeda, Nobukazu; Omori, Junji*; Nakahira, Masataka; Shibanuma, Kiyoshi
Journal of Nuclear Science and Technology, 41(12), p.1280 - 1286, 2004/12
The vibration experiments of the support structures with flexible plates for the ITER major components such as toroidal field coil and vacuum vessel were performed using small-sized flexible plates aiming to obtain its basic mechanical characteristics such as dependence of the stiffness on the loading angle. The experimental results obtained by the hammering and frequency sweep tests were agreed each other, so that the experimental method is found to be reliable. In addition, the experimental results were compared with the analytical ones in order to estimate an adequate analytical model. As a result, the bolt connection strongly affected on the stiffness of the support structure. After studies of modeling the bolts, it is found that the analytical results modeling the bolts with finite stiffness only in the axial direction and infinite stiffness in the other directions agree well with the experimental ones. Using this model, the stiffness of the support structure for the ITER major components can be calculated precisely in order to estimate the dynamic behaviors.
Takeda, Nobukazu; Shibanuma, Kiyoshi
Purazuma, Kaku Yugo Gakkai-Shi, 80(11), p.988 - 990, 2004/11
The simplified analytical model of the support structure composed of complicated structures such as multiple flexible plates was proposed for the dynamic analysis of the ITER major components of VV and TF coil. The support structure composed of flexible plates and connection bolts was modeled as a spring model composed of only two spring elements including the effect of connection bolts. The stiffness of both spring models for VV and TF coil agree well with that of shell models simulating actual structures such as flexible plates and connection bolts. Using the proposed model, the dynamic analysis of the VV and TF coil for the ITER were performed to estimate the integrity under the design earthquake at Rokkasho, a candidate of ITER site. As a result, it is found that the maximum relative displacement of 8.6 mm between VV and TF coil is much less than 100 mm, so that the integrity of the major components are ensured for the expected earthquake event.
Fujimoto, Nozomu; Tachibana, Yukio; Saikusa, Akio*; Shinozaki, Masayuki; Isozaki, Minoru; Iyoku, Tatsuo
Nuclear Engineering and Design, 233(1-3), p.273 - 281, 2004/10
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)From a viewpoint of heat leakage, there were two incidents during HTTR power-rise-tests. One was a temperature rise of the primary upper shielding, and the other was a temperature rise of the core support plate. Causes of the both incidents were small amount of helium flow in structures. For the temperature rise of the primary upper shielding, countermeasures to reduce the small amount of helium flow, enhancement of heat release and installation of thermal insulator were taken. For the temperature rise of the core support plate, temperature evaluations were carried out again considering the small amount of helium flow and design temperature of the core support plate was revised. By these countermeasures, the both temperatures were kept below their limits.
Sumita, Junya; Ishihara, Masahiro; Nakagawa, Shigeaki; Kikuchi, Takayuki; Iyoku, Tatsuo
Nuclear Engineering and Design, 233(1-3), p.81 - 88, 2004/10
Times Cited Count：4 Percentile：29.26(Nuclear Science & Technology)A High Temperature Gas-cooled Reactor is particularly attractive due to its capability of producing high temperature helium gas and its possibility to exploit inherent safety characteristic. To achieve high temperature helium-gas, reactor internals are made of graphite and heat resistant materials, its surroundings are composed of metals. The reactor internals of the HTTR consist of graphite and metallic core support structures and shielding blocks. This paper describes the reactor internal design of the HTTR, especially the core support graphite structures, and the program of an in-service inspection.
Sumita, Junya; Hanawa, Satoshi; Kikuchi, Takayuki; Ishihara, Masahiro
JAERI-Tech 2003-023, 37 Pages, 2003/03
JAERI-Tech-2003-023.pdf:7.79MB
Visual inspection of core support graphite structures using TV camera as in-service inspection and measurement of material characteristics using surveillance test specimens are planned in the High Temperature Engineering Test Reactor (HTTR) to confirm structural integrity of the core support graphite structures. For the visual inspection, in-service inspection system has been developed, and pre-service inspection using the system was carried out. As the result of the pre-service inspection, it was validated that high quality of visual inspection with TV camera can be carried out, and also structural integrity of the core support graphite structures at the initial stage of the HTTR operation was confirmed.
; Sato, Kazuyoshi; Araki, Masanori; Nakamura, Kazuyuki; Dairaku, Masayuki; ; Akiba, Masato
Fusion Technology, 30(3(PT.2A)), p.793 - 797, 1996/12
no abstracts in English
; Araki, Masanori; Nakamura, Kazuyuki; Sato, Kazuyoshi; ; Dairaku, Masayuki; Akiba, Masato
JAERI-Tech 95-033, 63 Pages, 1995/06
no abstracts in English
Nishio, Satoshi; *; *; *; *; Koizumi, Koichi; Abe, Tetsuya; *; Tada, Eisuke
JAERI-M 91-089, 138 Pages, 1991/05
no abstracts in English
Motoki, Yasuo; Hada, Kazuhiko; *; Baba, Osamu
JAERI-M 91-056, 44 Pages, 1991/03
no abstracts in English
Inagaki, Yoshiyuki; Fujimoto, Nozomu; Motoki, Yasuo; Iyoku, Tatsuo; Maruyama, So; Shiozawa, Shusaku
JAERI-M 90-223, 30 Pages, 1990/12
no abstracts in English
Inagaki, Yoshiyuki; Iyoku, Tatsuo; *; ; Shiozawa, Shusaku
JAERI-M 90-020, 70 Pages, 1990/02
no abstracts in English
;
Nucl.Eng.Des., 60, p.297 - 309, 1980/00
Times Cited Count：2 Percentile：33.85(Nuclear Science & Technology)no abstracts in English
