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Journal Articles

Outline of guideline for seismic response analysis method using 3D finite element model of reactor building

Choi, B.; Nishida, Akemi; Shiomi, Tadahiko; Kawata, Manabu; Li, Y.

Proceedings of 28th International Conference on Nuclear Engineering; Nuclear Energy the Future Zero Carbon Power (ICONE 28) (Internet), 7 Pages, 2021/08

In the seismic safety assessment of building structures in nuclear facilities, lumped mass models are conventionally used. However, they cannot possess the required high-accuracy evaluation of nuclear facilities, such as the local response at the equipment location in a reactor building. In this point of view, a seismic response analysis method using a three-dimensional finite element (3D FE) model is indispensable. Although, it has been reported that the analysis results obtained using 3D FE models vary greatly depending on the experience and knowledge of analysts, the quality of analysis results should be insured by developing a standard analysis method. In the Japan Atomic Energy Agency, we have developed a guideline for seismic response analysis methods that adopt 3D FE models of reactor buildings. The guideline consists of a main body, commentary, and several supplements; it also includes procedures, recommendations, points of attention, and a technical basis for conducting seismic response analysis using 3D FE models of reactor buildings. In this paper, the outline of the guideline and analysis examples based on the guideline are presented.

Journal Articles

Evaluation of the effects of differences in building models on the seismic response of a nuclear power plant structure

Choi, B.; Nishida, Akemi; Muramatsu, Ken*; Takada, Tsuyoshi*

Nihon Jishin Kogakkai Rombunshu (Internet), 20(2), p.2_1 - 2_16, 2020/02


no abstracts in English

Journal Articles

Failure behavior analyses of piping system under dynamic seismic loading

Udagawa, Makoto; Li, Y.; Nishida, Akemi; Nakamura, Izumi*

International Journal of Pressure Vessels and Piping, 167, p.2 - 10, 2018/11

 Times Cited Count:4 Percentile:50.81(Engineering, Multidisciplinary)

It is important to assure the structural Integrity of piping systems under severe earthquakes because those systems comprise the pressure boundary for coolant with high pressure and temperature. In this study, we examine the seismic safety capacity of piping systems under severe dynamic seismic loading using a series of dynamic-elastic-plastic analyses focusing on dynamic excitation experiments of 3D piping systems which was tested by NIED. Analytical results were consistent with experimental data in terms of natural frequency, natural vibration mode, response accelerations, elbow opening-closing displacements, strain histories, failure position, and low-cycle fatigue failure lives. Based on these results, we concluded that the analytical model used in the study can be applied to failure behavior evaluation for piping systems under severe dynamic seismic loading.

Journal Articles

Uncertainty evaluation of seismic response of a nuclear facility using simulated input ground motions

Choi, B.; Nishida, Akemi; Muramatsu, Ken*; Takada, Tsuyoshi*

Proceedings of 12th International Conference on Structural Safety & Reliability (ICOSSAR 2017) (USB Flash Drive), p.2206 - 2213, 2017/08

In order to clarify the influence of the difference of modeling method on the variation of the result of seismic response analysis of nuclear facility, seismic response analysis using various simulated input ground motions was carried out and the sensitivity analyses of the variations in seismic response was conducted. In particular, we focused on the maximum acceleration response of reactor building shear walls, the effect of modeling method on response result and the factors of response variation were described and discussed.

Journal Articles

Reliability enhancement of seismic risk assessment of NPP as risk management fundamentals; Quantifying epistemic uncertainty in fragility assessment using expert opinions and sensitivity analysis

Choi, B.; Nishida, Akemi; Itoi, Tatsuya*; Takada, Tsuyoshi*; Furuya, Osamu*; Muta, Hitoshi*; Muramatsu, Ken

Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 8 Pages, 2016/10

In this study, we address epistemic uncertainty in structure fragility estimation of nuclear power plants (NPPs). In order to identify and quantify dominant factors in fragility assessment, sensitivity analyses of seismic analysis results are conducted for a target NPP building using a three-dimensional finite element model and a conventional lumped mass model (embedded sway rocking model), and the uncertainty caused by the major factors is then evaluated. The results are used to classify epistemic uncertainty levels in a fragility estimation workflow for NPPs in several stages, and a graded knowledge tree technique, which can be used for future fragility estimations, is proposed.

Journal Articles

Confirmation of seismic integrity of HTTR against 2011 Great East Japan Earthquake

Ono, Masato; Iigaki, Kazuhiko; Shimazaki, Yosuke; Shimizu, Atsushi; Inoi, Hiroyuki; Tochio, Daisuke; Hamamoto, Shimpei; Nishihara, Tetsuo; Takada, Shoji; Sawa, Kazuhiro; et al.

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 12 Pages, 2016/06

On March 11th, 2011, the Great East Japan Earthquake of magnitude 9.0 occurred. When the great earthquake occurred, the HTTR had been stopped under the periodic inspection and maintenance of equipment and instrument. In the great earthquake, the maximum seismic acceleration observed at the HTTR exceeded the maximum value in seismic design. The visual inspection of HTTR facility was carried out for the seismic integrity conformation of HTTR. The seismic analysis was also carried out using the observed earthquake motion at HTTR site to confirm the integrity of HTTR. The concept of comprehensive integrity evaluation for the HTTR facility is divided into two parts. One is the inspection of equipment and instrument. The other is the seismic response analysis using the observed earthquake. For the basic inspections of equipment and instrument were performed for all them related to the operation of reactor. The integrity of the facilities is confirmed by comparing the inspection results or the numerical results with their evaluation criteria. As the result of inspection of equipment and instrument and seismic response analysis, it was judged that there was no problem to operate the reactor, because there was no damage and performance deterioration, which affects the reactor operation. The integrity of HTTR was also supported by the several operations without reactor power in cold conditions of HTTR in 2011, 2013 and 2015.

Journal Articles

Evaluation on seismic integrity of HTTR core components

Ono, Masato; Iigaki, Kazuhiko; Shimazaki, Yosuke; Tochio, Daisuke; Shimizu, Atsushi; Inoi, Hiroyuki; Hamamoto, Shimpei; Takada, Shoji; Sawa, Kazuhiro

Proceedings of International Topical Meeting on Research Reactor Fuel Management and Meeting of the International Group on Reactor Research (RRFM/IGORR 2016) (Internet), p.363 - 371, 2016/03

HTTR is graphite moderated and helium gas-cooled reactor with prismatic fuel elements and hexagonal blocks. Here, the graphite block is brittle materials and might be damaged by collision of neighboring blocks by the large earthquake. A seismic observation system is installed in the HTTR site to confirm a behavior of a seismic event. On March 11th, 2011, off the Pacific coast of Tohoku Earthquake of magnitude 9.0 occurred. After the accident at the TEPCO Fukushima Daiichi Nuclear Power Station, the safety of nuclear reactors is the highest importance. To confirm the seismic integrity of HTTR core components, the seismic analysis was carried out using the evaluation waves based on the relationship between the observed earthquake motion at HTTR site and frequency transfer function. In parallel, confirmation tests of primary cooling system on cold state and integrity confirmation of reactor buildings and component support structures were also carried out. As a result, it was found that a stress value of the graphite blocks satisfied an allowable value, and the integrity of the HTTR core components was ensured. The integrity of HTTR core components was also supported by the operation without reactor power in cold conditions of HTTR. The obtained data was compared with the normal plant data before the earthquake. As the result, the integrity of the HTTR facilities was confirmed.

JAEA Reports

Plan of vibration tests for estimation of seismic performance of ITER tokamak

Takeda, Nobukazu; Nakahira, Masataka

JAERI-Tech 2004-073, 59 Pages, 2005/01


The ITER toamak is composed of major components such as superconducting magnet and vacuum vessel whose operation temperatures are changed from room temperature to 4 K and room temperature to 200$$^{circ}$$C, respectively. The gravity support of the tokamak is flexible in order to accept the thermal deformation caused by temperature change. This structural feature causes the complex behaviors of the tokamak during seismic events. Therefore, the mechanical characteristics of the flexible support have to be investigated in detail. The present report describes the global plan of the series of vibration tests to estimate the seismic performance of the ITER tokamak. Although it is ideal that the vibration tests are carried out using a full-scale model, scale models are planned due to the limitation of the test facilities. The test results can be estimated by a scaling law. When the scaling law cannot be applied to some performances, the test is performed using a full-scale model. In addition, the other tests such as vacuum vessel and small-scaled models of the support structure are also planned.

JAEA Reports

Dynamic analysis of ITER tokamak based on results of vibration test using scaled model

Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka

JAERI-Tech 2004-072, 43 Pages, 2005/01


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.

Journal Articles

Seismic design

Iigaki, Kazuhiko; Hanawa, Satoshi

Nuclear Engineering and Design, 233(1-3), p.59 - 70, 2004/12

 Times Cited Count:3 Percentile:24.74(Nuclear Science & Technology)

The high temperature engineering test reactor (HTTR) was constructed on a sand layer formed during the Quaternary era. A seismometry system was installed in the HTTR facility in order to performed seismic analyses using a seismic observation record. The analysis model in the design was improved so that the simulation analysis result reproduced the seismic observation record. The dynamic analysis was carried out using an improvement model in order to compare seismic forces in the design. As the result, it was confirmed that the seismic forces obtained by the improvement model was approximately more conservative than the seismic force used in the design.

Journal Articles

Dynamic analysis of ITER tokamak using simplified model for support structure

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.

Journal Articles

Numerical evaluation of experimental models to investigate the dynamic behavior of the ITER tokamak assembly

Onozuka, Masanori*; Takeda, Nobukazu; Nakahira, Masataka; Shimizu, Katsusuke*; Nakamura, Tomomichi*

Fusion Engineering and Design, 69(1-4), p.757 - 762, 2003/09

 Times Cited Count:1 Percentile:11.39(Nuclear Science & Technology)

The dynamic behavior of the ITER tokamak assembly has been investigated. Three experimental models have been considered to validate the numerical analysis methods for the dynamic events, mainly seismic events. A 1/8-scaled tokamak model, which is based on the 1998 ITER design, is under construction. Non-linear vibration characteristics, such as damping, can only be identified by a full-scale model. Therefore, a full-scale gravity support structure for the coil system has been designed and will be tested. In addition, for the sub-scaled tokamak model, the VV is assumed to be a rigid structure. This assumption is to be verified using a 1/20-scaled model. The above experimental models and their testing conditions have analytically and numerically evaluated. For example, both the static and dynamic spring constants obtained by static analysis and eigen-value analysis, respectively, were evaluated to be in good agreement.

JAEA Reports

User's manual of SECOM2: A Computer code for seismic system reliability analysis

Uchiyama, Tomoaki; Oikawa, Tetsukuni; Kondo, Masaaki; Watanabe, Yuichi*; Tamura, Kazuo*

JAERI-Data/Code 2002-011, 205 Pages, 2002/03


This report is a user's manual of seismic system reliability analysis code SECOM2 developed at the JAERI for system reliability analysis, which is one of the tasks of seismic probabilistic safety assessment (PSA) of nuclear power plants (NPPs). The SECOM2 code has many functions such as calculation of component and system failure probabilities for given seismic motion levels at the site of an NPP based on the response factor method, calculation of accident sequence frequencies and the core damage frequency (CDF), importance analysis using various indicators, uncertainty analysis, and calculation of the CDF taking into account the effect of the correlations of responses and capacities of components. These analyses require the fault tree (FT) representing the occurrence condition of the system failures and core damage, information about responses and capacities of the components which compose the FT, and seismic hazard curve for the NPP site as input. This report presents calculation method used in the SECOM2 code and how to use those functions in the SECOM2 code.

JAEA Reports

User's manual for seismic analysis code "SONATINA-2V"

Hanawa, Satoshi; Iyoku, Tatsuo

JAERI-Data/Code 2001-021, 150 Pages, 2001/08


no abstracts in English

Journal Articles

Seismic study of High-Temperature Engineering Test Reactor core graphite structures

Iyoku, Tatsuo; Inagaki, Yoshiyuki; Shiozawa, Shusaku; *

Nuclear Technology, 99, p.158 - 168, 1992/08

 Times Cited Count:2 Percentile:28.1(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Seismic study on high temperature gas-cooled reactor core

JAERI 1322, 157 Pages, 1991/04


no abstracts in English

Journal Articles

Seismic response of high temperature gas-cooled reactor core with block-type fuel, IV; Seismic reaponse of actual core predicted from experimental and analytical results of two-dimensional core modela

; *

Nihon Genshiryoku Gakkai-Shi, 27(2), p.145 - 158, 1985/00

 Times Cited Count:2 Percentile:38.26(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

35 (Records 1-20 displayed on this page)