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, 12 Pages, 2024/00

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.

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.

A Simplified method for evaluating sloshing impact pressure on a flat roof based on Wagner's theory

Takaya, Shigeru; Fujisaki, Tatsuya*

Mechanical Engineering Journal (Internet), 7(3), p.19-00526_1 - 19-00526_10, 2020/06

https://doi.org/10.1299/mej.19-00526

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.

Development of prototype reactor maintenance, 3; Application to valves of sodium-cooled reactor prototype

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).

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

https://doi.org/10.1299/transjsme.17-00389

no abstracts in English

Numerical analysis of flow-induced vibration of large diameter pipe with short elbow

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.