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Yamamoto, Tomohiko; Watakabe, Tomoyoshi; Miyazaki, Masashi; Okamura, Shigeki; Miyagawa, Takayuki; Yokoi, Shinobu*; Fukasawa, Tsuyoshi*; Fujita, Satoshi*
Mechanical Engineering Journal (Internet), 11(2), p.23-00393_1 - 23-00393_21, 2024/04
Yada, Hiroki; Takaya, Shigeru; Morohoshi, Kyoichi*; Yokoi, Shinobu*; Miyagawa, Takayuki*
Mechanical Engineering Journal (Internet), 10(4), p.23-00044_1 - 23-00044_13, 2023/08
To develop rationalized maintenance plans for nuclear power plants, the characteristics of each plant must be considered. For sodium-cooled fast reactor (SFR) plants, constraints on inspections exist due to the specialty that equipment retaining sodium must be handled, which is one of the important points that must be considered in maintenance rationalization. In this study, we propose a maintenance optimization scheme, which is a design support tool, using risk information to develop a maintenance strategy based on the system based code (SBC) concept. The SBC concept intends to provide a theoretical procedure to optimize the reliability of structure, system and components (SSCs) by administrating every related engineering requirements throughout the life of the SSCs from design to decommissioning. ASME Boiler and Pressure Vessel Code, Code Case, N-875 was developed based on the SBC concept. The purpose of this study is to establish detailed procedures for the maintenance optimization scheme based on the procedure in Code Case N-875. Furthermore, a quantitative trial evaluation of the core support structure of the next SFR under development in Japan is also performed using the maintenance optimization scheme.
Fukasawa, Tsuyoshi*; Hirayama, Tomoyuki*; Yokoi, Shinobu*; Hirota, Akihiko*; Somaki, Takahiro*; Yukawa, Masaki*; Miyagawa, Takayuki; Uchita, Masato*; Yamamoto, Tomohiko; Miyazaki, Masashi; et al.
Nihon Kikai Gakkai Rombunshu (Internet), 89(924), p.23-00023_1 - 23-00023_17, 2023/08
no abstracts in English
Ikesue, Shunichi*; Morita, Hideyuki*; Sago, Hiromi*; Yokoi, Shinobu*; Yamamoto, Tomohiko
Proceedings of ASME 2023 Pressure Vessels and Piping Conference (PVP 2023) (Internet), 10 Pages, 2023/07
Yamamoto, Tomohiko; Watakabe, Tomoyoshi; Miyazaki, Masashi; Miyagawa, Takayuki*; Yokoi, Shinobu*; Okamura, Shigeki*; Fukasawa, Tsuyoshi*; Fujita, Satoshi*
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 7 Pages, 2023/05
Yada, Hiroki; Takaya, Shigeru; Morohoshi, Kyoichi*; Yokoi, Shinobu*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 9 Pages, 2022/08
In order to develop rationalized maintenance plans of nuclear power plants, it is necessary to consider characteristics of each plant. For sodium-cooled fast reactor (SFR) plants, there are constraints on inspections due to the specialty that sodium equipment needs to be handled, which is one of the important points when considering rationalization of maintenance. Therefore, we previously proposed a maintenance optimization scheme based on the System Based Code (SBC) concept. One of proposed scheme goals is to develop detailed procedures of preparing a rationalized maintenance plan. In this study, the procedures to determine inspections for potential degradation and additional inspections in terms of defense-in-depth have been further clarified. Furthermore, the modified maintenance optimization scheme is also illustrated by a quantitative trial evaluation of the core support structure of the next SFR under development in Japan.
Ikesue, Shunichi*; Morita, Hideyuki*; Ishii, Hidekazu*; Sago, Hiromi*; Yokoi, Shinobu*; Yamamoto, Tomohiko
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 10 Pages, 2022/07
Ikesue, Shunichi*; Morita, Hideyuki*; Ishii, Hidekazu*; Sago, Hiromi*; Yokoi, Shinobu*; Yamamoto, Tomohiko
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 8 Pages, 2021/07
In this paper, a new method is proposed for the nonlinear sloshing condition of a cylindrical tank, which can evaluate the vertical load acting on the roof and the horizontal load acting on the sidewall. This method is a combination of simplified equations for the liquid surface level and velocity proposed in the past study and the new pressure model modified from the existing model. A long calculation time as CFD analysis is not needed, because this method is consisted of simplified equations. The validity of this method was confirmed by comparing them with the CFD and the test. And future issues on the improvement of this method were clarified from the result.
Yokoi, Shinobu*; Kamishima, Yoshio*; Sadahiro, Daisuke*; Takaya, Shigeru
JAEA-Data/Code 2016-002, 38 Pages, 2016/07
Many efforts have been made to implement the System Based Code concept aiming at optimizing margins dispersed in existing codes and standards. Failure probability calculated based on statistical information such as a type of probability distribution, mean (or median) and variance (or standard deviation) for random variables is expected to be a promising quantitative index for margin optimization. Statistical information on material strength, which is also required to calculate the failure probability, has been already reported in JAEA-Data/Code 2015-002 "Structural Properties of Material Strength for Reliability Evaluation of Components of Fast Reactors -Austenitic Stainless Steels-" whereas others have not been identified yet. This report provides methodologies and basic ideas to determine statistical parameters of other random variables, especially variable loads, necessary for reliability evaluation.
Takaya, Shigeru; Watanabe, Daigo*; Yokoi, Shinobu*; Kamishima, Yoshio*; Kurisaka, Kenichi; Asayama, Tai
Journal of Pressure Vessel Technology, 137(5), p.051802_1 - 051802_7, 2015/10
Times Cited Count:2 Percentile:10.81(Engineering, Mechanical)The minimum wall thickness required to prevent seismic buckling of a reactor vessel in a fast reactor is derived using the System Based Code (SBC) concept. One of the key features of SBC concept is margin optimization; to implement this concept, the reliability design method is employed, and the target reliability for seismic buckling of the reactor vessel is derived from nuclear plant safety goals. Input data for reliability evaluation, such as distribution type, mean value, and standard deviation of random variables, are also prepared. Seismic hazard is considered to evaluate uncertainty of seismic load. Minimum wall thickness required to achieve the target reliability is evaluated, and is found to be less than that determined from a conventional deterministic design method. Furthermore, the influence of each random variable on the evaluation is investigated, and it is found that the seismic load has a significant impact.
Kato, Atsushi; Kubo, Shigenobu; Chikazawa, Yoshitaka; Hayafune, Hiroki; Yokoi, Shinobu*; Nakata, Shuhei*; Tani, Akihiro*; Shimakawa, Yoshio*
Proceedings of 2014 International Congress on the Advances in Nuclear Power Plants (ICAPP 2014) (CD-ROM), p.616 - 623, 2014/04
This paper focuses on loss of heat removal system (LOHRS) type event as Design Extension Condition (DEC) and describes candidates design measures to improve the decay heat removal system of JSFR against LOHRS type DEC. The design requirements are determined based on the Safety Design Criteria for Generation-IV Sodium-cooled fast reactor system. Effectiveness and reliability of the candidate design measures are discussed with preliminary evaluations.
Takaya, Shigeru; Watanabe, Daigo*; Yokoi, Shinobu*; Kamishima, Yoshio*; Kurisaka, Kenichi; Asayama, Tai
Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 6 Pages, 2013/07
In this paper, minimum wall thickness requirement of reactor vessel of fast reactor for seismic buckling is discussed on the basis of the System Based Code (SBC) concept. One of key concepts of SBC is the margin optimization. To implement this concept, reliability design method is employed, and the target reliability for seismic buckling of reactor vessel is derived from nuclear plant safety goals. Input data for reliability evaluation such as distribution type, mean value and standard deviation of random variable are prepared. Seismic hazard is considered to evaluate uncertainty of seismic load. Wall thickness needed to achieve the target reliability is evaluated, and as a result, it is shown that the minimum wall thickness can be reduced from that required by a deterministic design method.
Kamishima, Yoshio*; Yokoi, Shinobu*
JNC TJ9410 2005-002, 113 Pages, 2004/11
This research considered formation nature examination of an apparatus vertical isolation system, and building arrangement of an apparatus vertical isolation system, while conducting investigation about the pliability of the element design accompanying cycle change of a vertical isolation element, and the possibility of design rationalization.
Fukasawa, Tsuyoshi*; Hirayama, Tomoyuki*; Yokoi, Shinobu*; Hirota, Akihiko*; Somaki, Takahiro*; Yukawa, Masaki*; Miyagawa, Takayuki*; Uchita, Masato*; Yamamoto, Tomohiko; Miyazaki, Masashi; et al.
no journal, ,
The seismic integrity of sodium-cooled fast reactor (SFR) designs in nuclear power plants is of paramount importance. Based on the static loading test, this study investigates the force-displacement relationship and load transference in a three-dimensional seismic isolation system that is envisaged for use in reactor buildings. In SFR designs, the necessity for thin-walled structures to maintain high-temperature structure integrity can unintentionally compromise the seismic design. Consequently, addressing horizontal and vertical seismic forces become vital for ensuring seismic resilience. Currently, there are no specific codes or standards governing the integration of Three-dimensional seismic isolation systems into nuclear reactor buildings. However, current guidelines for the design of horizontal seismic isolation systems emphasize the necessity to clarify the force-displacement relationship and load transfer under conditions of superimposed horizontal and vertical loads. This study involves static loading tests performed on a half-scale specimen, which is subjected to horizontal and vertical loads exceeding the design basis ground motions for the SFR. The findings affirm that the system's horizontal supporting function maintains the segregation of horizontal and vertical load transference, even under seismic loads that exceed the design basis ground motions.
Yokoi, Shinobu*; Yamamoto, Tomohiko; Miyazaki, Masashi; Tanaka, Masaaki; Yamane, Yuma*; Nishiwaki, Yoshinori*; Sago, Hiromi*; Morita, Hideyuki*; Iwasaki, Akihisa*; Ikesue, Shunichi*
no journal, ,
The design basis ground motions have been revised to improve the seismic resistance of nuclear power plants. The reduction of seismic forces not only horizontally but also vertically has required more critical than in the past to ensure the seismic resistance of components. Notably, the design of a Sodium-Cooled Fast Reactor will require reducing the seismic forces applied to the components because of the components with thin wall thickness. To overcome this problem, the authors plan to introduce a seismic isolation system. When the sloshing wave height is small, it can be approximated with a linear vibration model. However, when the sloshing wave height increases and the sloshing becomes nonlinear, it is necessary to evaluate the wave height using other methods such as numerical analysis. Although the evaluation of nonlinear sloshing wave height is important, there are few examples which quantitatively evaluate the wave height of nonlinear sloshing. This paper reports on the development plan and an overview of the evaluation method for nonlinear sloshing wave height and load applied to cylindrical tanks.
Sago, Hiromi*; Yamamoto, Tomohiko; Miyazaki, Masashi; Tanaka, Masaaki; Yokoi, Shinobu*; Yamane, Yuma*; Nishiwaki, Yoshinori*; Morita, Hideyuki*; Iwasaki, Akihisa*; Ikesue, Shunichi*; et al.
no journal, ,
The design basis ground motions have been revised to improve the seismic resistance of nuclear power plants. The reduction of seismic forces not only horizontally but also vertically has required more critical than in the past to ensure the seismic resistance of components. Notably, the design of a Sodium-Cooled Fast Reactor will require reducing the seismic forces applied to the components because of the components with thin wall thickness. To overcome this problem, the authors plan to introduce a seismic isolation system. When the sloshing wave height is small, it can be approximated with a linear vibration model. However, when the sloshing wave height increases and the sloshing becomes nonlinear, it is necessary to evaluate the wave height using other methods such as numerical analysis. Although the evaluation of nonlinear sloshing wave height is important, there are few examples which quantitatively evaluate the wave height of nonlinear sloshing. This paper reports the results of the sloshing water test carried out to obtain test data for the construction of the evaluation method and the results of the reproduction analysis carried out using the VOF method.
Morita, Hideyuki*; Yamamoto, Tomohiko; Miyazaki, Masashi; Tanaka, Masaaki; Yokoi, Shinobu*; Sago, Hiromi*; Ikesue, Shunichi*
no journal, ,
The design basis ground motions have been revised to improve the seismic resistance of nuclear power plants. The reduction of seismic forces not only horizontally but also vertically has required more critical than in the past to ensure the seismic resistance of components. Notably, the design of a Sodium-Cooled Fast Reactor will require reducing the seismic forces applied to the components because of the components with thin wall thickness. To overcome this problem, the authors plan to introduce a seismic isolation system. When the sloshing wave height is small, it can be approximated with a linear vibration model. However, when the sloshing wave height increases and the sloshing becomes nonlinear, itis necessary to evaluate the wave height using other methods such as numerical analysis. Although the evaluation of nonlinear sloshing wave height is important, there are few examples which quantitatively evaluate the wave height of nonlinear sloshing. This paper reports on the study result of the predictive evaluation method for nonlinear sloshing wave height and impact load acting on the flat roof applied to cylindrical tanks.
Ikesue, Shunichi*; Yamamoto, Tomohiko; Miyazaki, Masashi; Tanaka, Masaaki; Yokoi, Shinobu*; Sago, Hiromi*; Morita, Hideyuki*
no journal, ,
The design basis ground motions have been revised to improve the seismic resistance of nuclear power plants. The reduction of seismic forces not only horizontally but also vertically has required more critical than in the past to ensure the seismic resistance of components. Notably, the design of a Sodium-Cooled Fast Reactor will require reducing the seismic forces applied to the components because of the components with thin wall thickness. To overcome this problem, the authors plan to introduce a seismic isolation system. However, the natural frequency of first order sloshing may be close to the response frequency of the Sodium-Cooled Fast Reactor with the seismic isolation system, and the sloshing wave height is expected to increase. When the sloshing wave height increases, the sloshing becomes the nonlinear sloshing, which can't be evaluated by linear sloshing theory. In order to evaluate the sloshing loads, which act on the roof and the internal structure, the nonlinear sloshing liquid surface shape and the nonlinear sloshing flow velocity are necessary. Therefore, the authors studied the predictive evaluation method of the nonlinear sloshing for the liquid surface shape and the flow velocity with simplified equations. This paper reports on an overview of this predictive evaluation method.
Yukawa, Masaki*; Morobishi, Ryota*; Yamamoto, Tomohiko; Hirayama, Tomoyuki*; Yokoi, Shinobu*
no journal, ,
A half-scale combined test specimen was fabricated for a three-dimensional seismic isolation device in which thicker laminated rubber bearing and disc spring units are arranged in series, and a static loading test was carried out by simultaneous loading in both the horizontal and vertical directions. The test results demonstrated that this three-dimensional seismic isolation device maintained the specified design performance and smooth operation, and the device concept was established.
Okamura, Shigeki*; Hirayama, Tomoyuki*; Yokoi, Shinobu*; Somaki, Takahiro*; Miyagawa, Takayuki*; Uchita, Masato*; Yamamoto, Tomohiko; Miyazaki, Masashi; Fukasawa, Tsuyoshi*; Fujita, Satoshi*
no journal, ,
no abstracts in English