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Ichihara, Yoshitaka*; Nakamura, Naohiro*; Nabeshima, Kunihiko*; Choi, B.; Nishida, Akemi
Nuclear Engineering and Design, 441, p.114160_1 - 114160_10, 2025/09
Times Cited Count：3 Percentile：73.73(Nuclear Science & Technology)This paper evaluates the applicability of equivalent linear analysis of reinforced concrete model, which uses frequency-independent complex damping with a small computational load, to the seismic design of nuclear power plant reactor buildings. To this end, a three-dimensional finite element method analysis of the soil-structure interaction focusing on nonlinear and equivalent linear seismic behavior of the building embedded in an ideally uniform soil condition was performed for the Kashiwazaki-Kariwa Nuclear Power Plant Unit 7 reactor building. The equivalent linear analysis results correlated well with the nonlinear analysis results of the shear strain, acceleration, displacement, and acceleration response spectrum, demonstrating the effectiveness of the equivalent linear analysis method. Moreover, the equivalent linear analysis results were more conservative than those of nonlinear analysis using the material constitutive law in evaluating the shear strain of the external wall of the reactor building. From this result, equivalent linear analysis method tended to obtain a lower building stiffness than nonlinear analysis under the analysis conditions used in this paper.
Choi, B.; Nishida, Akemi; Kawata, Manabu; Shiomi, Tadahiko; Li, Y.
JAEA-Research 2024-001, 206 Pages, 2024/03
JAEA-Research-2024-001.pdf:9.12MB
In the assessment of seismic safety and the design of building structures in nuclear facilities, lumped mass models have been used as standard methods. Recent advances in computer capabilities allow the use of three-dimensional finite element (3D FE) models to account for the 3D behavior of buildings, material nonlinearity, and the nonlinear soil-structure interaction effect. While 3D analysis method has many advantages, it is necessary to ensure its reliability as a new approach. The International Atomic Energy Agency performed an international benchmark study using the 3D FE analysis model for reactor building of Unit 7 at TEPCO's Kashiwazaki-Kariwa Nuclear Power Station based on recordings from the Niigataken Chuetsu-oki Earthquake in 2007. Multiple organizations from different countries participated in this study and the variation in their analytical results was significant, indicating an urgent need to improve the reliability of the analytical results by standardization of the analytical methods using 3D FE models. Additionally, it has been pointed out that it is necessary to understand the 3D behavior in the seismic fragility assessment of buildings and equipment, using realistic seismic response analysis method based on 3D FE models. In view of these considerations, a guideline for the seismic response analysis method using a 3D FE model was developed by incorporating the latest knowledge and findings in this area. The purpose of the guideline is to improve the reliability of the seismic response analysis method using 3D FE model of reactor buildings. The guideline consists of a main body, commentaries, and appendixes. The standard procedures, recommendations, key points to note, and technological bases for conducting seismic response analysis on reactor buildings using 3D FE models are provided in the guideline. In addition, the guideline will be revised reflecting the latest knowledge.
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.
Choi, B.; Nishida, Akemi; Shiomi, Tadahiko; Kawata, Manabu; Li, Y.
Transactions of the 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 10 Pages, 2022/07
In order to improve the seismic probabilistic risk assessment method, the authors are developing methods related to realistic response, realistic resistance and fragility assessment for buildings and equipment that are important for seismic safety. In this study, in order to identify of building damage mode subjected to large seismic motions, pushover analyses using multiple analysis codes were performed using a 3D FE model of a reactor building. We obtained the analysis results for the identification of local damage mode that contributes to the fragility assessment. In this paper, we report the progress of local damage mode and ultimate strength of the building by the pushover analysis. We also compared this result with the seismic response analysis results.
Ichihara, Yoshitaka*; Nakamura, Naohiro*; Nabeshima, Kunihiko*; Choi, B.; Nishida, Akemi
Kozo Kogaku Rombunshu, B, 68B, p.271 - 283, 2022/04
This paper aims to evaluate the applicability of the equivalent linear analysis method for reinforced concrete, which uses frequency-independent hysteretic damping, to the seismic design of reactor building of the nuclear power plant. To achieve this, we performed three-dimensional FEM analyses of the soil-structure interaction system, focusing on the nonlinear and equivalent linear seismic behavior of a reactor building under an ideal soil condition. From these results, the method of equivalent analysis showed generally good correspondence with the method of the nonlinear analysis, confirming the effectiveness. Moreover, the method tended to lower the structural stiffness compared to the nonlinear analysis model. Therefore, in the evaluation of the maximum shear strain, we consider that the results were more likely to be higher than the results of nonlinear analysis.
Choi, B.; Nishida, Akemi; Kawata, Manabu; Shiomi, Tadahiko; Li, Y.
JAEA-Research 2021-017, 174 Pages, 2022/03
JAEA-Research-2021-017.pdf:9.33MB
Standard methods such as lumped mass models have been used in the assessment of seismic safety and the design of building structures in nuclear facilities. Recent advances in computer capabilities allow the use of three-dimensional finite element (3D FE) models to account for the 3D behavior of buildings, material nonlinearity, and the nonlinear soil-structure interaction effect. Since the 3D FE model enables more complex and high-level treatment than ever before, it is necessary to ensure the reliability of the analytical results generated by the 3D FE model. Guidelines for assuring the dependability of modeling techniques and the treatment of nonlinear aspects of material properties have already been created and technical certifications have been awarded in domains other than nuclear engineering. The International Atomic Energy Agency performed an international benchmark study in nuclear engineering. Multiple organizations reported on the results of seismic response studies using the 3D FE model based on recordings from the Niigata-ken Chuetsuoki Earthquake in 2007. The variation in their analytical results was significant, indicating an urgent need to improve the reliability of the analytical results by standardization of the analytical methods using 3D FE models. Additionally, it has been pointed out that it is necessary to understand the 3D behavior in the seismic fragility assessment of buildings and equipment, which requires evaluating the realistic nonlinear behavior of building facilities when assessing their seismic fragility. In view of these considerations, a standard guideline for the seismic response analysis method using a 3D FE model was produced by incorporating the latest knowledge and findings in this area. The purpose of the guideline is to improve the reliability of the seismic response analysis method using 3D FE model of reactor buildings. The guideline consists of a main body, commentaries, and appendixes; it also provides standard procedures
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.
Ioka, Ikuo; Kuriki, Yoshiro*; Iwatsuki, Jin; Kubo, Shinji; Katsuyama, Jinya; Inagaki, Yoshiyuki
Mechanical Engineering Journal (Internet), 3(3), p.15-00619_1 - 15-00619_8, 2016/06
The IS process for hydrogen production has been developed by JAEA as application of a high-temperature gas cooled reactor. The IS process includes a severe corrosion environment which is made to boil and decompose concentrated sulfuric acid. The two-layer pipe consisted of the Fe-high Si alloy with boiling sulfuric acid-resistant and the carbon steel with the ductility was produced by centrifugal casting. The evaluation of characteristics was carried out. The Fe-high Si alloy lining showed enough corrosion resistance in boiling sulfuric acid. As evaluation of the Fe / Fe-high Si alloy interface, thermal cycle test was executed. There was no detachment of the interface though the cracks were generated in the vicinity of the interface. It is believed that the cracks parallel to the interface is attributed to the tensile stress during the thermal cyclic test using FEM analysis and the flake graphite precipitate. It was confirmed that the interface possessed the enough strength.
Shibata, Katsuyuki*; Onizawa, Kunio; Suzuki, Masahide; Li, Y.*
Nihon Kikai Gakkai M&M 2005 Zairyo Rikigaku Kanfarensu Koen Rombunshu, p.299 - 300, 2005/11
no abstracts in English
Tsuji, Nobumasa*; Shibata, Taiju; Sumita, Junya; Ishihara, Masahiro; Iyoku, Tatsuo
Advances in Fracture and Strength, Part 1, p.2720 - 2726, 2005/00
Times Cited Count：0 Percentile：0.00(Materials Science, Ceramics)The core restraint mechanisms have been made of steel materials in existing HTGR, but steel materials cannot keep tying force in higher coolant temperature over 5000
C caused by rapid stress relaxation. The C/C composite is one of the attractive candidates for core restraint mechanisms because of its excellent mechanical properties in high temperature. In the present study, basic design of the core restraint mechanism using C/C composite was conducted and structural integrity of C/C composite was evaluated by FEM stress analysis with anisotropic material model. As a result, it was shown that the core restraint mechanism using C/C composite has realizable possibility to achieve high temperature coolant up to 10000C in HTGRs.
Suzuki, Motoe; Saito, Hiroaki*
JAERI-Data/Code 2003-019, 423 Pages, 2003/12
JAERI-Data-Code-2003-019.pdf:17.7MB
A light water reactor fuel analysis code FEMAXI-6 is an advanced version which has been produced by integrating the former version with a number of improvements. In particular, the FEMAXI-6 code has attained a complete coupled solution of thermal analysis and mechanical analysis, permitting an accurate prediction of pellet-clad gap size and PCMI in high burnup fuel rods. Also, such new models have been implemented as pellet-clad bonding and fission gas bubble swelling, and the coupling with burning analysis code has been enhanced. Furthermore, a number of new materials properties and parameters have been introduced. With these advancements, the FEMAXI-6 code is a versatile tool not only in the normal operation but also in transient conditions. This report describes the design, basic theory, models and numerical method, improvements, and model modification. In order to facilitate effective and wide-ranging application of the code, formats and methods of input/output, and a sample output in an actual form are included.
Futakawa, Masatoshi; Wakui, Takashi*; Kogawa, Hiroyuki; Date, Hidefumi*
Proceedings of JSME/ASME Interntional Conference on Materials and Processing 2002, p.558 - 563, 2002/10
This paper describes the evaluation method of mechanical-property from micro or nanoscopic views, using the indentation technique. The depth sensing indentation technique was applied to evaluate the changes of mechanical properties due to the liquid metal corrosion. The effect of imposed stresses on the corrosion was taken by the residual stress developed around a large indent into account. An inverse analysis was applied to quantitatively evaluate the material degradation on the interface in contact with the liquid metal. This technique is useful to evaluate the material degradation of a sub-surface of materials such as coatings and corroded layer as well as a small area such as a crack tip, etc.
Suzuki, Motoe
JAERI-Data/Code 2000-030, 280 Pages, 2000/09
JAERI-Data-Code-2000-030.pdf:11.06MB
no abstracts in English
Suzuki, Motoe
Nuclear Engineering and Design, 201(1), p.99 - 106, 2000/09
Times Cited Count：7 Percentile：45.63(Nuclear Science & Technology)no abstracts in English
Hashimoto, Kazuichiro; Nakamura, Naohiko; Igarashi, Minoru*; Maruyama, Yu; Sugimoto, Jun
The 3rd JSME/ASME Joint Int. Conf. on Nuclear Engineering, Vol. 3, 0, p.1241 - 1246, 1995/00
no abstracts in English
I.Smid*; Akiba, Masato; Araki, Masanori; ; Sato, Kazuyoshi
JAERI-M 93-149, 200 Pages, 1993/07
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
JAERI-M 91-129, 347 Pages, 1991/08
no abstracts in English
Koizumi, Koichi; Hasegawa, Mitsuru*; Yoshida, Kiyoshi
JAERI-M 91-121, 200 Pages, 1991/08
no abstracts in English
JAERI-M 91-061, 119 Pages, 1991/04
no abstracts in English
