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Journal Articles

Stress intensity factor solutions for circumferential through-wall cracks applicable to pool type sodium cooled fast reactors

Yada, Hiroki; Takaya, Shigeru; Machida, Hideo*

Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 7 Pages, 2024/08

Journal Articles

Development of ARKADIA for the innovation of advanced nuclear reactor design process (Development status of the design optimization support tool, ARKADIA-Design)

Tanaka, Masaaki; Doda, Norihiro; Hamase, Erina; Kuwagaki, Kazuki; Mori, Takero; Okajima, Satoshi; Kikuchi, Norihiro; Yoshimura, Kazuo; Matsushita, Kentaro; Hashidate, Ryuta; et al.

Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2024/06

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. In this paper, focusing on the ARKADIA-Design, achievements in the development of optimization processes in the fields of the core design, the plant structure design, and the maintenance schedule planning, as major function of ARKADIA-Design, and numerical analysis methods to be used for the detailed analysis to confirm the plant performance after optimization are introduced at this point in time.

Journal Articles

Proposal of a method of extracting design issues on maintenance for the rational design of next-generation reactors

Hashidate, Ryuta; Yada, Hiroki; Takaya, Shigeru; Enuma, Yasuhiro

Hozengaku, 23(1), p.103 - 111, 2024/04

In this paper, we propose a new method for extracting design issues on maintenance. Maintenance periods might be prolonged due to design issues. In the proposed method, maintenance preconditions are extracted by organizing the design information. A maintenance schedule is created by using extracted maintenance preconditions. If the created maintenance schedule doesn't achieve target periods, design issues could be extracted from the viewpoint of maintenance precondition. A simple example using Monju design information is presented to illustrate the proposed method.

Journal Articles

Comparison between fracture mechanics evaluation methods in ASME Boiler & Pressure Vessel Code, section XI and those in JSME leak-before-break evaluation guidelines for sodium-cooled fast reactors

Yada, Hiroki; Takaya, Shigeru; Machida, Hideo*

Proceedings of ASME 2023 Pressure Vessels and Piping Conference (PVP 2023) (Internet), 8 Pages, 2023/09

ASME Boiler and Pressure Vessel code (BPVC), Section XI, Division 2 provides requirements for protecting passive components that affect reliability of the plant. It generally consists of technology-neutral common requirements, and additional ones for individual reactor types. Currently, an Appendix for sodium-cooled fast reactors (SFRs) is being developed based on Code Case N-875. In the Code Case, continuous leakage monitoring was employed as inspection method for components retaining liquid sodium. It is also important to introduce leak-before-break (LBB) assessment procedures in the Appendix because demonstration of LBB is necessary to show the adequacy of applying continuous leakage monitoring to the component of interest. However, LBB assessment method is not provided in ASME BPVCs. On the other hand, recently, LBB assessment guidelines for SFRs has been developed by the Japan Society of Mechanical Engineers (JSME). It could be used to prepare LBB assessment procedures for the Appendix, but it needs to confirm the consistency with ASME BPVC Sec. XI. In this study, fracture evaluation methods for pipes with through-wall crack are compared between JSME LBB assessment guidelines and applicable evaluation method in ASME BPVC Sec. XI, Div. 1.

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

Development of ARKADIA for the innovation of advanced nuclear reactor design process (Overview of optimization process development in design optimization support tool, ARKADIA-Design)

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.

Journal Articles

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

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.

Journal Articles

Development of numerical analysis codes for multi-level and multi-physics approaches in an advanced reactor design study

Tanaka, Masaaki; Doda, Norihiro; Mori, Takero; Yokoyama, Kenji; Uwaba, Tomoyuki; Okajima, Satoshi; Matsushita, Kentaro; Hashidate, Ryuta; Yada, Hiroki

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

Japan Atomic Energy Agency is developing an innovative design system named ARKADIA to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In the first phase of its development, ARKADIA-Design for design study and ARKADIA-Safety for safety assessment will be developed individually. In this paper, focusing on the ARKADIA-Design, the concept of the system is described and numerical analysis codes to be used for the multi-level and multi-physics analyses are introduced. Descriptions of the practical functions composed by the analysis codes and the representative problems in application studies for validation are introduced.

Journal Articles

Fundamental study on scheduling of inspection process for fast reactor plants

Suzuki, Masaaki*; Ito, Mari*; Hashidate, Ryuta; Takahashi, Keita; Yada, Hiroki; Takaya, Shigeru

2020 9th International Congress on Advanced Applied Informatics (IIAI-AAI 2020), p.797 - 801, 2021/07

Journal Articles

Development of leak before break assessment guidelines for sodium cooled fast reactors in Japan

Yada, Hiroki; Wakai, Takashi; Miyagawa, Takayuki*; Machida, Hideo*

Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 10 Pages, 2021/07

Journal Articles

Proposal of maintenance rationalization for next-generation fast reactors based on the analysis of the prolonged maintenance of the prototype fast-breeder reactor in Japan, "Monju", 1; Analysis of plant schedule of "Monju" in cold shutdown

Hashidate, Ryuta; Toyota, Kodai; Takahashi, Keita; Yada, Hiroki; Takaya, Shigeru

Hozengaku, 19(4), p.115 - 122, 2021/01

In order to improve both safety and economic efficiency of a nuclear power plant, it is necessary to realize rational maintenance based on characteristics of the plant. The prototype fast-breeder reactor in Japan, Monju, spent most of the year for the maintenance. Thus, it is important to identify causes of the prolonged maintenance of Monju and to investigate countermeasures for implementation of rational maintenance of next-generation fast reactors. In this study, the authors investigated the causes of the prolonged maintenance of Monju during reactor cold shutdown based on the plant schedule of Monju. In addition, we proposed the maintenance optimization idea for next-generation fast reactors to solve the revealed issues.

Journal Articles

Proposal of inspection rationalization method and application for sodium cooled fast reactor

Yada, Hiroki; Takaya, Shigeru; Enuma, Yasuhiro

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08

Journal Articles

Development of in-service inspection rules for liquid-metal cooled reactors using the system based code concept

Takaya, Shigeru; Asayama, Tai; Yada, Hiroki; Roberts, A. T.*; Schaaf, F.*

Journal of Pressure Vessel Technology, 142(2), p.021601_1 - 021601_5, 2020/04

 Times Cited Count:1 Percentile:6.67(Engineering, Mechanical)

Inservice inspection rules for liquid-metal cooled plants were historically provided by Section XI, Division 3 of the ASME Boiler and Pressure Vessel Code. However, some parts of the Code remained as being in the course of preparation. Although no major revisions were made to Division 3 since the first issue in 1980, a newly developed and published Code Case N-875, now provides alternative examinations to the methods previously contained in Division 3. The Code Case was developed using the System Based Code concept pursuing rationalization of codes and standards based on reliability targets throughout a plant's service life. In this paper, an overview of the Code Case is presented. The technical foundation to establish the applicability of these alternative examinations as delineated in the Code Case, consists of Stage I and II evaluations with compensating individual considerations. Stage I is a structural integrity evaluation without the contribution of inservice inspections, while Stage II is evaluation of the detectability of a postulated flaw. Not only conventional direct detection methods, but also indirect detection methods are permitted to be employed through the Stage II evaluation. Furthermore, the detailed evaluation procedures are illustrated through the application of the Code Case's evaluation criteria to the primary heat transport piping system of a prototype sodium-cooled fast breeder reactor in Japan, specifically Monju.

Journal Articles

Development of LORL evaluation method and its application to a loop-type sodium-cooled fast reactor

Imaizumi, Yuya; Yamada, Fumiaki; Arikawa, Mitsuhiro*; Yada, Hiroki; Fukano, Yoshitaka

Mechanical Engineering Journal (Internet), 5(4), p.18-00083_1 - 18-00083_11, 2018/08

A calculation program was developed to evaluate and discuss the effectiveness of the countermeasures such as sodium pump-up and siphon-breaking against the loss-of-reactor-level (LORL) where the coolant circulation path is lost in loop-type sodium-cooled fast reactors. Due to the non-negligible possibility obtained by probabilistic risk assessment (PRA), sodium leakages in two points both occurred in primary heat transport system (PHTS) was assumed in this study. In addition, the crack size was discussed and evaluated realistically, instead of the value that was assumed in the conventional studies. Representative sequences and leakage positions were chosen, and the sodium level transient in reactor vessel (RV) was calculated. The calculations were also conducted where the larger crack size was set for the second leakage, in order to investigate additional requirements to maintain the RV sodium level. The evaluation results clarified that the coolant circulation loop can be maintained even after the second leakage in PHTS, taking into account the effects by the countermeasures.

Journal Articles

Effect of 3-D initial imperfections on the deformation behaviors of head plates subjected to convex side pressure

Yada, Hiroki; Ando, Masanori; Tsukimori, Kazuyuki; Ichimiya, Masakazu*; Anoda, Yoshinari*

Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 9 Pages, 2018/07

Containment vessel (CV) of nuclear power plants is an important structure to prevent radioactive release, however, the safety margin of the CV against pressure are not numerically clarified. The head plate structure is included in CV boundary of fast reactor. In order to develop the evaluation method of the ultimate strength of the head plate structure at beyond the specified limit, pressure failure tests and finite element analysis (FEA) of the head plates subjected to convex side pressure were performed. In the test of the relative thin thickness head plate, non-axisymmetric deformations was observed in post buckling behavior and failure pressure was lower than other cases. To evaluate non-axisymmetric deformations in the test, FEA using 3-D solid model constructed by precise dimensions of the test specimen, moreover, FEA using simplified model with uniform or non-uniform thickness were performed. Through analyses, the feature of the post buckling behavior was discussed.

Journal Articles

Leak rate tests of penetrate cracked head plates and modeling of head plate thickness distribution for 3-D analyses

Tsukimori, Kazuyuki*; Yada, Hiroki; Ando, Masanori; Ichimiya, Masakazu*; Anoda, Yoshinari*

Proceedings of 12th International Conference on Asian Structure Integrity of Nuclear Components (ASINCO-12) (CD-ROM), p.105 - 121, 2018/04

In FBR plants the head plate constitutes a part of the boundary of the containment vessel (CV), therefore, it is an important issue if the function as the boundary is maintained or not in the severe accident. And also it is important to evaluate the leak rate from the penetrated crack of the head plate, in order to estimate the effect of released fission product out of CV. Authors conducted pressure endurance tests of head plate specimens subjected to external pressure, which covered post-buckling behaviors and until crack penetration. In this paper leak rate test results at several pressure levels are introduced and the tendency of leak rate behaviors with relation of the penetrate crack length and the pressure level are discussed. Also, the modeling of head plate thickness distribution for 3-D analyses based on the detailed 3-D measurement data of specimens is discussed, which possibly relates to the 3-D deformation patterns observed in the tests and the length of penetration cracks.

Journal Articles

Proposal on LBB evaluation conditions for sodium cooled fast reactor pipes and effects of pipe parameters

Yada, Hiroki; Takaya, Shigeru; Wakai, Takashi; Nakai, Satoru; Machida, Hideo*

Nihon Kikai Gakkai Rombunshu (Internet), 84(859), p.17-00389_1 - 17-00389_15, 2018/03

no abstracts in English

Journal Articles

Experimental study on the deformation and failure of the bellows structure beyond the designed internal pressure

Ando, Masanori; Yada, Hiroki; Tsukimori, Kazuyuki; Ichimiya, Masakazu*; Anoda, Yoshinari*

Journal of Pressure Vessel Technology, 139(6), p.061201_1 - 061201_12, 2017/08

 Times Cited Count:1 Percentile:6.93(Engineering, Mechanical)

In this study, in order to develop the evaluation method of the ultimate pressure of the bellows structure subject to the internal pressure beyond the specified, the failure test and finite element analysis (FEA) of the bellows structure were performed. The failure modes were demonstrated through the series of tests, and three kind of failure mode were observed. To simulate the buckling and deformation behavior during the test, the implicit and explicit analyses were performed.

103 (Records 1-20 displayed on this page)