Development of a surrogate system of a plant dynamics simulation model and an abnormal situation identification system for nuclear power plants using deep neural networks

Seki, Akiyuki; Yoshikawa, Masanori; Nishinomiya, Ryota*; Okita, Shoichiro; Takaya, Shigeru; Yan, X.

Nuclear Technology, 210(6), p.1003 - 1014, 2024/06

https://doi.org/10.1080/00295450.2023.2273566

Two types of deep neural network (DNN) systems have been constructed with the intent to assist safety operation of a nuclear power plant. One is a surrogate system (SS) that can estimate physical quantities of a nuclear power plant in a computational time of several orders less than a physical simulation model. The other is an abnormal situation identification system (ASIS) that can estimate the state of the disturbance causing an anomaly from physical quantities of a nuclear power plant. Both systems are trained and tested using data obtained from the analytical code for incore and plant dynamics (ACCORD), which reproduces the steady and dynamic behavior of the actual high Temperature engineering test reactor (HTTR) under various scenarios. The DNN models are built by adjusting, the main hyperparameters. Through these procedures, these systems are shown able to perform with a high degree of accuracy.

Proposal of a novel AI-based plant operator support system for the safety of nuclear power plants

Takaya, Shigeru; Seki, Akiyuki; Yoshikawa, Masanori; Sasaki, Naoto*; Yan, X.

Mechanical Engineering Journal (Internet), 11(2), p.23-00408_1 - 23-00408_11, 2024/04

https://doi.org/10.1299/mej.23-00408

Enhancing the ability to manage abnormal situations is important for improvement of the safety of nuclear power plants. It is necessary to investigate potential risks thoroughly in advance, and prepare countermeasures against the identified risks. In case of an occurrence of an abnormal situation, plant operators are required to recognize the plant situation promptly and select a suitable countermeasure. This study develops a novel plant operator support system designed not only to estimate details of anomalies in a plant but also propose countermeasures adaptively by employing several AI technologies of deep neural network and reinforcement learning. The design and performance of the proposed system is illustrated using High Temperature engineering Test Reactor operated in Japan Atomic Energy Agency.

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.

Attention-based time series analysis for data-driven anomaly detection in nuclear power plants

Dong, F.*; Chen, S.*; Demachi, Kazuyuki*; Yoshikawa, Masanori; Seki, Akiyuki; Takaya, Shigeru

Nuclear Engineering and Design, 404, p.112161_1 - 112161_15, 2023/04

https://doi.org/10.1016/j.nucengdes.2023.112161

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

https://doi.org/10.1299/jsmepes.2022.26.A221

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.

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

https://doi.org/10.1109/IIAI-AAI50415.2020.00157

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.

Influence of dead weight and internal pressure to seismic buckling probability of fast reactor vessels

Takaya, Shigeru; Sasaki, Naoto*

Mechanical Engineering Journal (Internet), 7(3), p.19-00549_1 - 19-00549_9, 2020/06

https://doi.org/10.1299/mej.19-00549

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.

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

https://doi.org/10.1115/1.4044344

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.

Proposal of a simple evaluation method for sloshing impact pressure on flat roofs

Takaya, Shigeru; Fujisaki, Tatsuya*

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 6 Pages, 2018/07

https://doi.org/10.1115/ICONE26-82562

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.