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Journal Articles

Proposal for maintenance optimization scheme based on system based code concept

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.

Journal Articles

Proposal of detailed procedures of determining rational in-service inspection requirements based on system based code concept

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.

Journal Articles

Study on the predictive evaluation method for loads acting on roof and sidewall of cylindrical tank in nonlinear sloshing based on simplified equations

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.

JAEA Reports

Determination methodologies for input data including loads considered for reliability evaluation of fast reactor components

Yokoi, Shinobu*; Kamishima, Yoshio*; Sadahiro, Daisuke*; Takaya, Shigeru

JAEA-Data/Code 2016-002, 38 Pages, 2016/07

JAEA-Data-Code-2016-002.pdf:1.51MB

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.

Journal Articles

Study on minimum wall thickness requirement for seismic buckling of reactor vessel based on system based code concept

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:11.65(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.

Journal Articles

Design approach for decay heat removal systems based on the safety design criteria for Gen-IV sodium-cooled fast reactor

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.

Journal Articles

Study on minimum wall thickness requirement of reactor vessel of fast reactor for seismic buckling by system based code

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.

JAEA Reports

The Investigation about the impact of the vibration period change in the vertical isolation element

Kamishima, Yoshio*; Yokoi, Shinobu*

JNC TJ9410 2005-002, 113 Pages, 2004/11

JNC-TJ9410-2005-002.pdf:5.84MB

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.

Oral presentation

Development of a reliability evaluation method for fatigue-creep interaction failure of an FBR reactor vessel

Takaya, Shigeru; Okajima, Satoshi; Asayama, Tai; Chitose, Hiromasa*; Machida, Hideo*; Yokoi, Shinobu*; Kamishima, Yoshio*

no journal, , 

An evaluation method of the occurrence probability of a through-wall crack in a reactor vessel of a fast breeder reactor due to fatigue-creep interaction has been proposed. Input data were prepared for a trial evaluation and the proposed evaluation method was applied. The result was compared with the allowable occurrence probability derived from the safety requirements for FBR.

Oral presentation

Study on event progression in PLOHS for JSFR, 1; Numerical simulation of plant dynamics phase in PLOHS

Matsuo, Eiji*; Watanabe, Motoko*; Yamada, Yumi*; Koyama, Kazuya*; Shimakawa, Yoshio*; Sato, Mitsuru*; Kamishima, Yoshio*; Yokoi, Shinobu*; Yamano, Hidemasa; Suzuki, Toru; et al.

no journal, , 

Event transition in PLOHS is investigated to contribute to Level 2 PSA focused on JSFR. This report will describes about the evaluation result of the event escalation in the response transition after PLOHS.

Oral presentation

Design approach for LOHRS type event based on the safety design criteria for Generation-IV SFR, 2; Study of equipment specification

Kato, Atsushi; Kubo, Shigenobu; Chikazawa, Yoshitaka; Nakata, Shuhei*; Yokoi, Shinobu*

no journal, , 

Plant design studies against "Loss of heat removal system" event were carried out based on the safety design criteria for 4th generation sodium-cooled fast reactor discussing in the Generation-IV international forum (GIF-SFR-SDC). This presentation introduces equipment specification study to match the GIF-SFR-SDC.

Oral presentation

Development of design support tools (Automatic and optimized design of the expansion tank)

Yokoi, Shinobu*; Okafuji, Takashi*; Onoda, Yuichi; Tanaka, Masaaki

no journal, , 

no abstracts in English

Oral presentation

Research and development of three-dimensional isolation system for SFR (Concept of isolation system and development plan)

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

Oral presentation

Development of three-dimensional seismic isolation system for sodium-cooled fast reactor

Yamamoto, Tomohiko; Miyazaki, Masashi; Watakabe, Tomoyoshi; Miyagawa, Takayuki*; Uchita, Masato*; Yokoi, Shinobu*; Somaki, Takahiro*

no journal, , 

no abstracts in English

Oral presentation

Impact resistance evaluation for CDA in a pool type reactor with common specifications

Eto, Masao*; Yokoi, Shinobu*; Yamashita, Masato*; Miura, Kazuhiro*; Okafuji, Takashi*; Onoda, Yuichi; Yamamoto, Tomohiko; Uchita, Masato*

no journal, , 

no abstracts in English

Oral presentation

Research and development of three-dimensional isolation system; Static loading test for beyond design basis conditions using half scale model

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.

Oral presentation

Study on the predictive evaluation method of nonlinear sloshing wave height and load of cylindrical tanks, 1; Development plan

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.

Oral presentation

Study on the predictive evaluation method of nonlinear sloshing wave height and load of cylindrical tanks, 2; Shaking table test and analysis for nonlinear sloshing

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.

Oral presentation

Study on the predictive evaluation method of nonlinear sloshing wave height and load of cylindrical tanks, 3; Validation of the predictive evaluation method for nonlinear sloshing wave height and impact load acting on flat roof

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.

Oral presentation

Study on the predictive evaluation method of nonlinear sloshing wave height and load of cylindrical tanks, 4; Study on nonlinear sloshing wave height and flow velocity

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.

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