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Dong, F.*; Chen, S.*; Demachi, Kazuyuki*; Yoshikawa, Masanori; Seki, Akiyuki; Takaya, Shigeru
Nuclear Engineering and Design, 404, p.112161_1 - 112161_15, 2023/04
Times Cited Count：0Yada, 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.
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
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
Takaya, Shigeru; Fujisaki, Tatsuya*
Mechanical Engineering Journal (Internet), 7(3), p.19-00526_1 - 19-00526_10, 2020/06
In severe seismic conditions, sloshing waves are considered to even reach a roof slab of a reactor vessel. The structural integrity of roof slabs is required to be evaluated against sloshing impacts. However, there is no widely recognized evaluation method for sloshing impact pressure on flat roofs yet. Therefore, in this paper, a simplified evaluation method is proposed based on Wagner's theory, which is a well-known classic theory for evaluating impact pressures on rigid wedges dropping on water surfaces. In the proposed method, we assume an equivalent wedge on a flat roof. The impact pressure on the equivalent wedge is evaluated by applying Wagner's theory. Computational fluid dynamics analysis is conducted to confirm that a key assumption of Wagner's theory is applicable to the evaluation of sloshing impact on a flat roof. In addition, the predictability of the proposed method is investigated by comparing literature data of sloshing experiments with the estimated values.
Takaya, Shigeru; Sasaki, Naoto*
Mechanical Engineering Journal (Internet), 7(3), p.19-00549_1 - 19-00549_9, 2020/06
Seismic buckling of vessels is one of main concerns for the design of fast reactor plants in Japan. In previous studies, we discussed evaluation methods of seismic buckling probability of vessels by taking account of seismic hazards in order to rationalize seismic buckling evaluation, and proposed a rule for seismic buckling of vessels based on the load and resistant factor design method. The proposed method deals with only seismic load, but in actuality, dead weight and internal pressure also exist. In this study, the rule was expanded so that dead weight and internal pressure can be taken into account. Furthermore, the influences of dead weight and internal pressure to seismic buckling evaluation were discussed. As result, it was shown that approximately 10 to 20% of further rationalization of allowable seismic load could be achieved by considering dead weight and internal pressure in the evaluation.
Chikazawa, Yoshitaka; Takaya, Shigeru; Tagawa, Akihiro; Kubo, Shigenobu
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 6 Pages, 2019/05
A maintenance management required to prototype nuclear power reactors has been developed. One of important mission of a prototype reactor is to develop maintenance program for commercial reactors step by step securing safety. Since operating experience at the early stage is limited, the maintenance program for the prototype reactor should be a progressive one. It has to be modified and improved frequently taking into account R&D insight and operation experiences. Additionally, the maintenance program has to consider features of the prototype reactor even at the early stage. To select maintenance grades on particular components/systems, risk informed and graded approaches are effective. And maintenance programs have to take into account degradation mechanism originally due to reactor features. In this paper, applications for maintenance program on sodium valves of prototype fast breeder reactor Monju are studied as an example of prototype sodium-cooled reactors (SFR).
Takaya, Shigeru; Fujisaki, Tatsuya*
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 6 Pages, 2018/07
Sloshing is one of important issues for both loop-type and tank-type fast reactors with free liquid surface. Periods of seismic vibration are lengthened by base isolation systems installed to prevent damages to facilities during earthquakes, and get close to the natural periods of sloshing. As a result, sloshing is promoted. Sloshing waves are assumed to even reach a roof slab of a reactor vessel in severe seismic conditions. It is important to evaluate structural integrity for sloshing impacts on roofs. However, there are not any established evaluation methods for impact pressure on flat roofs yet. Therefore, in this study, a simple evaluation method is proposed based on Wagner's theory. The effectiveness of the proposed method is illustrated using computational fluid dynamics analysis and literature data of sloshing experiments.
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; Takaya, Shigeru
Nuclear Engineering and Design, 323, p.463 - 473, 2017/11
Times Cited Count：2 Percentile：22.85(Nuclear Science & Technology)In the present study, a method for creep-fatigue life evaluation of Mod.9Cr-1Mo steel weld joint was proposed based on finite element analysis (FEA). Since the point of the creep-fatigue life evaluation in the weld joint is a consideration of the metallurgical discontinuities, FEA was performed using a model with three material properties, a base metal (BM), weld metal (WM) and a heat-affected zone (HAZ) formed in the base metal due to the welding heat input, to consider the mutual relationships among them. The material properties of these three materials were collected and utilized in FEA for considering such metallurgical discontinuities. The creep-fatigue life estimated using the proposed evaluation method based on the FEA results were compared with available creep-fatigue test data, and the proposed method was found to predict the number of cycles to failure within a factor of 3.
Takaya, Shigeru; Fujisaki, Tatsuya*; Tanaka, Masaaki
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 8 Pages, 2017/07
Flow-induced vibration (FIV) of a hot-leg piping is one of main concerns in the design of an advanced loop-type sodium cooled fast reactor. We have been developing numerical analysis models to deal with this issue. In this study, computational fluid dynamics (CFD) simulation of a 1/3 scaled-model of the hot-leg piping was conducted. The results such as velocity profiles and power spectral densities (PSD) of pressure fluctuations were compared with experiment ones. The simulated PSD of pressure fluctuation at the recirculation region agreed well with the experiment. Then, stress induced by FIV was evaluated using pressure fluctuation data calculated by the CFD simulation. The calculated stress generally agrees well the measurement values, which indicates the importance of precise evaluation of the PSD of pressure fluctuation at the recirculation region for evaluation of FIV of the hot-leg piping with a short elbow.
Takaya, Shigeru; Sasaki, Naoto*; Asayama, Tai; Kamishima, Yoshio*
Mechanical Engineering Journal (Internet), 4(3), p.16-00558_1 - 16-00558_12, 2017/06
In this study, we developed a new design rule for the prevention of seismic buckling of vessels using the load and resistance factor design method to enable more rational vessel designs. The effectiveness of the new design rule was illustrated in comparison with the current provision.
Kotake, Shoji*; Chikazawa, Yoshitaka; Takaya, Shigeru; Otaka, Masahiko; Kubo, Shigenobu; Arai, Masanobu; Kunogi, Kosuke; Ito, Takaya*; Yamaguchi, Akira*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
A maintenance management required to prototype nuclear power reactors is proposed. Monitoring and control of sodium impurity and thermal transient are extremely important for sodium boundary maintenance for sodium-cooled fast reactors. At the fast stage of the prototype reactor Monju operation, degradation mechanism on the piping should be demonstrated based on operation experiences. Therefore inspection on a representative position for crack indication and pipe thickness is proposed. Due to less experience of SFR plants, early detection of boundary failure is considered. For a matured operation stage, when degradation mechanism is well demonstrated based on inspection data, inspection cycle could be extended. And for commercial reactors, maintenance without inspection will be established based on accumulated operation experiences including those of the prototype reactor Monju.
Arai, Masanobu; Kunogi, Kosuke; Aizawa, Kosuke; Chikazawa, Yoshitaka; Takaya, Shigeru; Kubo, Shigenobu; Kotake, Shoji*; Ito, Takaya*; Yamaguchi, Akira*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
Applications for maintenance program on piping support of prototype fast breeder reactor Monju are studied. Based on degradation mechanism, snubbers in Monju primary cooling system showed lifetime more than the plant lifetime of 30 years by experiments conservatively. For the first step during construction, visual inspection on accessible all supports could be available. In that visual inspection, mounting conditions and damages of all accessible supports could be monitored. One of major features of the Monju primary piping system is large thermal expansion due to large temperature difference between maintenance and operation conditions. Thanks to that large thermal expansion, integrity of the piping support could be monitored by measuring piping displacement. When technologies of piping displacement monitoring are matured in Monju, visual inspection on piping support could be shifted to piping displacement monitoring. At that stage, the visual inspection could be limited only on representative supports.
Takaya, Shigeru; Chikazawa, Yoshitaka; Hayashida, Kiichi; Tagawa, Akihiro; Kubo, Shigenobu; Yamashita, Atsushi
Hozengaku, 15(4), p.71 - 78, 2017/01
A maintenance management suitable to nuclear power plants (NPP) at R&D stage was discussed. Objectives of maintenance management of NPP at R&D stage was first clarified. Next, applicability of codes for maintenance management of commercial NPP to NPP at R&D stage was discussed. Then, requirements and consideration for maintenance management of NPP at R&D stage was proposed. Finally, an example that the proposal was applied to setting maintenance program of sodium-cooled fast reactor was presented.
Okajima, Satoshi; Takaya, Shigeru; Asayama, Tai
Nihon Kikai Gakkai Rombunshu (Internet), 83(845), p.16-00434_1 - 16-00434_13, 2017/01
no abstracts in English
Ando, Masanori; Takaya, Shigeru
Nuclear Engineering and Design, 310, p.217 - 230, 2016/12
Times Cited Count：1 Percentile：12.01(Nuclear Science & Technology)In the present study, to develop an evaluation procedure and design rules for Mod.9Cr-1Mo steel weld joints, a method for creep-fatigue life evaluation of Mod.9Cr-1Mo steel weld joints was proposed based on the finite element analysis (FEA) results with a series of cyclic plate bending tests of the longitudinal and horizontal seamed plates. The strain concentration and redistribution behaviors were evaluated and failure cycles were estimated using FEA by considering the test conditions and metallurgical discontinuities in the weld joints. The elastic follow-up factors calculated from a comparison of the elastic and inelastic FEA results were determined to be less than 1.5. Based on the estimated elastic follow-up factors obtained via inelastic FEA, a simplified technique using elastic FEA was proposed for evaluating the creep-fatigue life in Mod.9Cr-1Mo steel weld joints.
