Seismic fragility evaluation of a thin-walled vessel in a sodium-cooled fast reactor

Watakabe, Tomoyoshi; Yamano, Hidemasa; Futagami, Satoshi

Transactions of the 28th International Conference on Structural Mechanics in Reactor Technology (SMiRT28) (Internet), 10 Pages, 2025/08

The fragility evaluation method of the RV in the SFR has been developed for both failure mode buckling and fatigue. The past studies showed the fragility using the RV of a loop-type SFR: radius R / thickness t is about 100. Recently, Japan Atomic Energy Agency together with Japanese industrial partners has been conducting a pool-type SFR design. The fragility of the pool-type RV has never been evaluated by the proposed method, which is based on the fatigue failure mode instead of the buckling. To investigate the basic behaviour of the pool-type RV under excessive seismic loading, this report investigated the fragility by using a simple cylindrical model with a wall thickness close to the design of the pool-type RV: R/t is 200, as the first trial.

Effectiveness evaluation of the measures for improving resilience of nuclear structures against excessive earthquake, 1; Fragility evaluation of reactor vessel based on structural analysis

Nishino, Hiroyuki; Kurisaka, Kenichi; Futagami, Satoshi; Watakabe, Tomoyoshi; Yamano, Hidemasa

Proceedings of Probabilistic Safety Assessment and Management & Asian Symposium on Risk Assessment and Management (PSAM17 & ASRAM2024) (Internet), 10 Pages, 2024/10

The reactor vessel (RV) buckling was a dominant contributor to core damage. However, even if the RV is buckled due to seismic shaking, it is expected that the RV maintains stable state without unstable failure such as rupture, collapse. Realistic consideration of the post-buckling behavior is regarded as a measure for improving the resilience in this study. The purpose of this study is to understand the post-buckling deformation behavior of the RV and to evaluate the RV fragility based on fatigue failure. This study performed structural analysis using a finite element method to quantify time histories of displacement, strain, etc. As the result of the analysis, wrinkles of the buckling appeared at the elevation higher than the liquid level in the RV. The largest strain value was also indicated around this elevation. The cumulative fatigue damage fraction was evaluated in this analysis to evaluate the fragility of fatigue failure in addition to the buckling fragility. The result showed that the seismic intensity for the median fragility of the fatigue failure was about six times larger than the design-basis ground motion. This is 1.2 times larger than the buckling-based result, which suggests that realistic evaluation of the post-buckling behavior could contribute to improving the resilience of the nuclear structure.

Research and development of three-dimensional isolation system for sodium-cooled fast reactor, 8; Assembly static test results of three-dimensional isolated device by bi-axial loadings

Somaki, Takahiro*; Yukawa, Masaki*; Fukasawa, Tsuyoshi*; Hirayama, Tomoyuki*; Uchita, Masato*; Miyagawa, Takayuki; Okamura, Shigeki; Yamamoto, Tomohiko; Watakabe, Tomoyoshi; Miyazaki, Masashi; et al.

Proceedings of the ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 9 Pages, 2024/07

https://doi.org/10.1115/PVP2024-124495

Research and development of three-dimensional isolation system for sodium cooled fast reactor, 7; Development summary of three-dimensional isolation system

Watakabe, Tomoyoshi; Yamamoto, Tomohiko; Okamura, Shigeki; Miyazaki, Masashi; Miyagawa, Takayuki; Uchita, Masato*; Hirayama, Tomoyuki*; Somaki, Takahiro*; Yukawa, Masaki*; Fukasawa, Tsuyoshi*; et al.

Proceedings of the ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 10 Pages, 2024/07

To secure the seismic safety of the thin-walled mechanical components and piping under a severe design earthquake level, employing a three-dimensional (3D) seismic isolation system has been planned in a sodium-cooled fast reactor. The development results of the 3D isolation system have been reported in previous papers so far. Its update is reported in Part 7 to Part 9. Part 7 describes the overview of the development, the test plan of the isolation system in the assembled state of each element, and the performance of individual isolation elements. In part 8, the performance of the isolation device that each element was assembled into was investigated through loading tests. Part 9 reports analytical studies by an analysis model validated based on the insight of the test results.

Seismic qualification of crossover piping systems on a sodium-cooled fast reactor with a seismic isolation system

Watakabe, Tomoyoshi; Okuda, Takahiro; Okajima, Satoshi

Mechanical Engineering Journal (Internet), 11(2), p.23-00395_1 - 23-00395_13, 2024/04

https://doi.org/10.1299/mej.23-00395

A three-dimensional seismic isolation system is planed for application to the conceptual design of a sodium-cooled fast reactor (SFR) in Japan. The crossover piping is laid between the nuclear building with the isolation system and the turbine building without the isolation system. A large displacement of the nuclear building with the isolation system is imposed on the crossover piping, which situation is a particular seismic issue because of the isolation system employment. Furthermore, it should be considered that the SFR operates at elevated temperatures compared with light water reactors. In this study, seismic evaluation using an example of a crossover piping layout was performed in accordance with the elevated temperature code of Japan Society of Mechanical Engineers. According to the evaluation results and the up to date technologies such as knowledge obtained from existing dynamic failure tests of piping components, an appropriate seismic evaluation method for the crossover piping was studied.

Upgrade of seismic design procedure for piping systems based on elastic-plastic response analysis

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

https://doi.org/10.60316/jcossar.10.0_143

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.

Effect of the plasticity of pipe and support on the seismic response of piping systems

Okuda, Takahiro; Takahashi, Hideki*; Watakabe, Tomoyoshi

Mechanical Engineering Journal (Internet), 10(4), p.23-00075_1 - 23-00075_9, 2023/08

https://doi.org/10.1299/mej.23-00075

In recent years, to make the seismic design more rational for the piping systems in nuclear power plants, it has been expected to develop a design method considering plastic deformation and the accompanying energy dissipation of the piping itself. In this study, an extensive series of seismic response analyses was conducted to investigate the degree of influence of the plastic deformation of the pipe support structures on the seismic response of the entire piping system. The analyses include; plasticity is considered for (1) none, (2) the piping only, (3) the support structure only, and (4) both the piping and the support structure.

Seismic evaluation of crossover piping on a sodium cooled fast reactor with three-dimensional isolation system

Watakabe, Tomoyoshi; Okuda, Takahiro; Okajima, Satoshi

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 5 Pages, 2023/05

A three-dimensional seismic isolation system is planed for application to the conceptual design of a sodium-cooled fast reactor (SFR) in Japan. The crossover piping is laid between the nuclear building with the isolation system and the turbine building without the isolation system. A large displacement of the nuclear building with the isolation system is imposed on the crossover piping, which situation is a particular seismic issue because of the isolation system employment. Furthermore, it should be considered that the SFR operates at elevated temperatures compared with light water reactors. In this study, seismic evaluation using an example of a crossover piping layout was performed in accordance with the elevated temperature code of Japan Society of Mechanical Engineers. According to the evaluation results and the up to date technologies such as knowledge obtained from existing dynamic failure tests of piping components, an appropriate seismic evaluation method for the crossover piping was studied.

Sodium-cooled Fast Reactors

Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.

Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07

https://shop.elsevier.com/books/sodium-cooled-fast-reactors/morishita/978-0-12-824076-2

This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.

A JSME code case on piping seismic design based on inelastic response analysis and strain-based fatigue criteria

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

https://doi.org/10.1115/1.4045674

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.