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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.
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.
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.
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.
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.
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.
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
Toyota, Kodai; Hashidate, Ryuta; Takahashi, Keita; Yada, Hiroki; Takaya, Shigeru
Hozengaku, 20(2), p.95 - 103, 2021/07
Yada, Hiroki; Wakai, Takashi; Miyagawa, Takayuki*; Machida, Hideo*
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 10 Pages, 2021/07
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.
Yada, Hiroki; Takaya, Shigeru; Enuma, Yasuhiro
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08
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.
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.
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.
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
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：7.39(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.
Tsukimori, Kazuyuki; Ando, Masanori; Yada, Hiroki; Ichimiya, Masakazu*; Anoda, Yoshinari*; Arakawa, Manabu*
Transactions of the 24th International Conference on Structural Mechanics in Reactor Technology (SMiRT-24) (USB Flash Drive), 10 Pages, 2017/08
The analytical treatment of Multi-ply bellows behaviours is difficult compared with that of single-ply bellows, since the uncertainty of friction between plies exists. In this study verification was conducted based on experiments by comparing between two-ply and single-ply bellows test results. Following results were obtained. The spring rate of two-ply bellows is approximately twice of that of single-ply bellows, even if internal pressure is loaded. Typical buckling behaviour of bellows, in-plane squirm, was observed in both cases of two-ply and single-ply bellows. The deformation patterns were similar with each other, but the pressure levels of two-ply bellows were approximately twice of those of single-ply bellows. These means the friction force can be ignored practically. As the conclusion, two-ply bellows analyses can be replaced by the analyses of single-ply bellows model with half pressure load and the effort of numerical analysis can be reduced.
Ando, Masanori; Yada, Hiroki; Tsukimori, Kazuyuki; Ichimiya, Masakazu*; Anoda, Yoshinari*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 11 Pages, 2017/07
In this study, in order to develop the evaluation method of the pressure toughness of bellows structure under the beyond design base event, the pressure failure tests and finite element analysis (FEA) of the bellows structure subjected to internal pressure were performed. In the test of five convolutions 0.5 mm-thickness bellows specimen with guard pipe, the maximum pressure was larger than those in the tests without guard pipe specimens and ductile failure was observed. On the other hand, in the test of five convolutions 0.5 mm-thickness bellows specimen without guard pipe, local failure was observed. In the test of the six convolutions 1.0 mm-thickness bellows specimen, ductile failure was observed in the both single and double ply bellows. The maximum pressure obtained in all tests were about 10 times larger than the estimated results of limiting design pressure based on in-plain instability by the EJMA standards.
Yada, Hiroki; Ando, Masanori; Tsukimori, Kazuyuki; Ichimiya, Masakazu*; Anoda, Yoshinari*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 8 Pages, 2017/07
The head plate that composes the boundary between primary and secondary coolant in intermediate heat exchanger of FBR has an important role when the progress of the BDBE is considered. In order to develop the evaluation method of the pressure toughness of the head plate under the BDBE, the pressure failure tests and finite element analysis of two types of head plate subjected to convex side pressure was performed in this study. It can be concluded that a failure mode of a head plate subjected convex side pressure is circumferential through-wall crack caused by straightening following the bending deformation near the rim.
Yada, Hiroki
Hozengaku, 15(3), P. 86, 2016/10
no abstracts in English
