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Okajima, Satoshi; Mori, Takero; Kikuchi, Norihiro; Tanaka, Masaaki; Miyazaki, Masashi
Mechanical Engineering Journal (Internet), 10(4), p.23-00042_1 - 23-00042_12, 2023/08
In this paper, we propose the simplified procedure to estimate failure probability of components subjected to thermal transient for the design optimization. Failure probability can be commonly used as an indicator of component integrity for various failure mechanisms. In order to reduce number of analyses required for one estimation, we have adopted the First Order Second Moment (FOSM) method as the estimation method of failure probability on the process of the optimization, and an orthogonal table in experiment design method is utilized to define conditions of the analyses for the evaluation of the input parameters for the FOSM method. The superposition of ramp responses is also utilized to evaluate the time history of thermal transient stress instead of finite element analysis. Through the demonstration study to optimize thickness of cylindrical vessel subjected to thermal transient derived from shutdown, we confirmed that the procedure can evaluate the failure probability depending on the cylinder thickness with practical calculation cost.
Okajima, Satoshi; Mori, Takero; Kikuchi, Norihiro; Tanaka, Masaaki; Miyazaki, Masashi
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 7 Pages, 2022/08
In this paper, we propose the simplified procedure to estimate failure probability of components subjected to thermal transient for the design optimization. Failure probability can be commonly used as an indicator of component integrity for various failure mechanisms. In order to reduce number of analyses required for one estimation, we have adopted the First Order Second Moment (FOSM) method as the estimation method of failure probability on the process of the optimization, and an orthogonal table in experiment design method is utilized to define conditions of the analyses for the evaluation of the input parameters for the FOSM method. Through the demonstration study to optimize thickness of cylindrical vessel subjected to thermal transient derived from shutdown, we confirmed that the procedure can evaluate the failure probability depending on the cylinder thickness with practical calculation cost.
Kikuchi, Norihiro; Mori, Takero; Okajima, Satoshi; Tanaka, Masaaki; Miyazaki, Masashi
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07
In JAEA, the design optimization method for plant structure has been developed on the process to output optimal solution of the thickness of reactor vessel wall against thermal transient and seismic loads in a SFR as a representative problem. Resistance characteristic of the wall on the load derived from thermal transient is one of the most important factors for safety estimation on the structural integrity. Failure probability of components against thermal transient was set to one of variables in the objective function for a common scale to compare with other variables in different failure mechanisms. In the iterative process to achieve the optimal solution, a number of evaluations to measure the influence on the load derived from thermal transient was necessarily conducted. More reduction of required time for evaluations is desired. To perform the iterative evaluation process efficiently, the automatization of parametric analyses was implemented in the optimization process.
Tanaka, Masaaki; Doda, Norihiro; Yokoyama, Kenji; Mori, Takero; Okajima, Satoshi; Hashidate, Ryuta; Yada, Hiroki; Oki, Shigeo; Miyazaki, Masashi; Takaya, Shigeru
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07
To assist conceptual studies of various reactor systems conducted by private sectors in nuclear power innovation, development of an innovative design system named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) is undergoing to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In this paper, focusing on the ARKADIA-Design as a part of it, the progress in the development of optimization processes on the representative problems in the fields of the core design, the plant structure design, and the maintenance schedule planning are introduced.
Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Ando, Masanori; Miyazaki, Masashi
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 9 Pages, 2021/07
Buckling evaluation methods capable of evaluating elasto-plastic buckling under axial compression, bending, and shear loads are required for cylindrical vessels of fast reactors to cope with thinning due to increasing diameter and application to the seismic isolation design against huge seismic ground motion. In this study, in order to confirm the applicability of the proposal evaluation method, several buckling tests and FE analyses were carried out using the specimens made of Modified 9Cr-1Mo steel. The buckling modes and strength data in the load region where the interaction of cyclic axial compression, bending and shear buckling could occur were examined. As a result, it was confirmed that the proposal evaluation method estimated the buckling load in the tests conservatively. Moreover, a series of finite element analyzes using a model with residual stress due to welding revealed that the effect of residual stress on buckling strength is negligible in the evaluation method.
Fukasawa, Tsuyoshi*; Miyagawa, Takayuki*; Uchita, Masato*; Yamamoto, Tomohiko; Miyazaki, Masashi; Okamura, Shigeki*; Fujita, Satoshi*
Nihon Kikai Gakkai Rombunshu (Internet), 87(898), p.21-00007_1 - 21-00007_17, 2021/06
This paper describes a fundamental study on the seismic safety margin for the isolated structure using laminated rubber bearings. The variation of the seismic response assumed in the isolated structure will occur under the superposition of "Variations in seismic response due to input ground motions" and "Error with design value accompanying manufacture of the isolation devices ". The seismic response analysis which allows to their conditions is important to assess the seismic safety margin for the isolated structure. This paper clarifies that the seismic safety margin of the isolated structure, which consists of rubber bearings, for Sodium-cooled Fast Reactor (SFR) is ensured against the basis ground motions of Japan Electric Association Guide 4601 (JEAG4601) and SFR through the seismic response analysis considering the variation factors of seismic response. In addition, a relationship between the seismic safety margin and the excess probability of linearity limits is discussed using the results of seismic response analysis.
Miyazaki, Yasunori; Sano, Yuichi; Okamura, Nobuo; Watanabe, Masayuki; Koka, Masashi*
QST-M-29; QST Takasaki Annual Report 2019, P. 72, 2021/03
no abstracts in English
Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Ando, Masanori; Miyazaki, Masashi
Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 9 Pages, 2020/08
Buckling evaluation methods capable of evaluating elasto-plastic buckling under axial compression, bending, and shear loads are required for cylindrical vessels of fast reactors to cope with thinning due to increasing diameter and application to the seismic isolation design against huge seismic ground motion. In this study, in order to confirm the applicability of the proposal evaluation method, several buckling tests and FE analyses were carried out using the specimens made of Modified 9Cr-1Mo steel. The buckling modes and strength data in the load region where the interaction of axial compression, bending and shear buckling could occur were examined. As a result, it was confirmed that the proposal evaluation method estimated the buckling load in the tests conservatively. In addition, buckling strength evaluated by elasto-plastic buckling analysis had good accuracy compared to each test result by considering the stress-strain relationship and imperfection of test specimen.
Ohara, Takaharu; Ikeda, Makinori; Morishita, Masaki; Miyazaki, Masashi; Shimada, Koji; Moriizumi, Makoto*
no journal, ,
no abstracts in English
Ishimaru, Tsuneari; Shimada, Koji; Sasaki, Akimichi; Tanaka, Yukumo; Miyazaki, Masashi; Yasue, Kenichi; Niwa, Masakazu; Sueoka, Shigeru; Umeda, Koji; Ikeda, Makinori
no journal, ,
The Nuclear Regulation Authority instructed to develop a further additional research plan regarding the crushed zones on the Monju site against the JAEA on September 25. The JAEA developed a research plan in response to this instruction, and, the stripping investigation with expansion of the area, the additional detailed geographical and geological surveys around the mountains/terrace boundary, and the marine seismic surveys in the coastal waters, were conducted. The fracture zones were grouped into 2 systems called 
-system and 
-system in the on-site investigation. We examined the cross-cutting relationships and displacements of the fracture zones and confirmed that the -system was formed after the -system. From the investigation results so far, similar to the results of the compilation report of April 31, 2010, no clear evidence was observed to support that on-site crushed zones are active faults.
Shimada, Koji; Tateishi, Ryo*; Ishimaru, Tsuneari; Sasaki, Akimichi; Tanaka, Yukumo; Miyazaki, Masashi; Yasue, Kenichi; Niwa, Masakazu; Sueoka, Shigeru; Umeda, Koji; et al.
no journal, ,
Activity evaluation of crush zones encountered in basement rock is an issue of the seismic safety assessment of nuclear plant and geological isolation of radioactive wastes. The selection of crush zone of which has been evaluated should be defined as the latest one by means of turn determination of crush zone activity based on stratigraphic or structural geological method. A lesson from granitic basement rock (Kojaku granite) holding the fast breeder reactor "Monju" is presented. The Kojaku Granite form the oval Tsuruga peninsula (ca. 8km in width) on the southeastern coast of the Sea of Japan and the age is 68.5 plus/minus 0.7Ma (Zircon U-Pb age).1. Stratigraphy-oriented turn determination of crush zone activity. (1.1) Turn determination using cover sediments. (1.2) Turn determination using dyke, mineral and clay vein. 2. Structural-oriented turn determination of crush zone activity.
Miyazaki, Yasunori; Watanabe, So; Sano, Yuichi; Kofuji, Hirohide; Takeuchi, Masayuki; Ishigami, Ryoya*; Koka, Masashi*; Sato, Takahiro*
no journal, ,
no abstracts in English
-ray resistant assessment of MA adsorbent for the next generation vitrificationMiyazaki, Yasunori; Sano, Yuichi; Kofuji, Hirohide; Watanabe, So; Koka, Masashi*; Sato, Takahiro*
no journal, ,
no abstracts in English
Miyazaki, Yasunori; Sano, Yuichi; Kofuji, Hirohide; Watanabe, So; Koka, Masashi*; Sato, Takahiro*
no journal, ,
no abstracts in English
Okuda, Takahiro; Miyazaki, Masashi; Kitamura, Seiji
no journal, ,
The equipment and piping in nuclear power plants receive loads from various directions during earthquakes. Understanding the load direction that most affects the structural strength of piping is an important for evaluating the seismic integrity of piping. In this paper, focusing on thin wall tee applied in fast reactor, a result of analytical evaluation of impact on structural strength of a thin wall tee is reported.
Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Ando, Masanori; Miyazaki, Masashi
no journal, ,
Buckling evaluation methods capable of evaluating elasto-plastic buckling under axial compression, bending, and shear loads are required for cylindrical vessels of fast reactors to cope with thinning due to increasing diameter and application to the seismic isolation design against huge seismic ground motion. In this study, in order to cnfirm the applicability of the proposal evaluation method, several buckling tests and FE analyses were carried out using the specimens made of Modified 9Cr-1Mo steel. The buckling modes and strength data in the load region where the interaction of axial compression, bending and shear buckling could occur were examined. As a result, it was confirmed that the proposal evaluation method estimated the buckling load in the tests conservatively. In addition, buckling strength evaluated by elasto-plastic buckling analysis had good accuracy compared to each test result by considering the stress-strain relationship and imperfection of test specimen.
Tanaka, Masaaki; Oki, Shigeo; Miyazaki, Masashi; Takaya, Shigeru; Yokoyama, Kenji; Ando, Masanori
no journal, ,
Design optimization support tool named ARKADIA-Design for diverse nuclear power plant has been developing in Japan Atomic Energy Agency. Its development plans, goals, and schedules in the fields of core, plant structure including thermal-hydraulics, and maintenance are presented with brief introduction of progress at present.
Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Ando, Masanori; Miyazaki, Masashi
no journal, ,
Buckling evaluation methods capable of evaluating elasto-plastic buckling under axial compression, bending, and shear loads are required for cylindrical vessels of fast reactors to cope with thinning due to increasing diameter and application to the seismic isolation design against huge seismic ground motion. In this study, the effect of initial imperfection and stress of large vessels manufactured by welding multiple plates on the buckling load was evaluated via a series of analyses. In these analyses, the superimposition of global imperfection over the entire vessel due to accumulation of misalignment in buckling mode imperfection had minimal effect on the buckling load. Accordingly, the length for measuring the imperfection amplitude should correspond to the wavelength of the buckling mode. The local imperfection and residual stress by welding also exerted small effects on the buckling load.
Tanaka, Masaaki; Doda, Norihiro; Yokoyama, Kenji; Mori, Takero; Okajima, Satoshi; Hashidate, Ryuta; Yada, Hiroki; Oki, Shigeo; Miyazaki, Masashi; Takaya, Shigeru
no journal, ,
Design optimization support tool named ARKADIA-Design for diverse nuclear power plant is being developed in Japan Atomic Energy Agency. To implement the design optimizations in the fields of core, plant structure, and maintenance, development of the numerical analysis methods including a coupled analysis to evaluate the plant performance and installation of the procedure in each optimization process are investigated.
Miyagawa, Takayuki*; Yamamoto, Tomohiko; Miyazaki, Masashi; Okamura, Shigeki*; Fukasawa, Tsuyoshi*; Fujita, Satoshi*; Hirayama, Tomoyuki*
no journal, ,
no abstracts in English
