CNN-based acoustic identification of gas-liquid jet; Evaluation of noise resistance and visual explanation using Grad-CAM

Mikami, Nao*; Ueki, Yoshitaka*; Shibahara, Masahiko*; Aizawa, Kosuke; Ara, Kuniaki*

International Journal of Multiphase Flow, 171, p.104688_1 - 104688_13, 2024/01

https://doi.org/10.1016/j.ijmultiphaseflow.2023.104688

State sensing of bubble jet flow based on acoustic identification of deep learning

Mikami, Nao*; Ueki, Yoshitaka*; Shibahara, Masahiko*; Aizawa, Kosuke; Ara, Kuniaki

Proceedings of 17th International Heat Transfer Conference (IHTC-17) (Internet), 9 Pages, 2023/08

Study on performance evaluation of self-actuated shutdown system for sodium-cooled fast reactor; Investigation on flow field around curie point electromagnet

Aizawa, Kosuke; Hiyama, Tomoyuki; Kobayashi, Jun; Kurihara, Akikazu

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 6 Pages, 2023/05

Self-actuated shutdown system (SASS) is a passive reactor-shutdown system that utilizes a Curie-point electromagnet (CPEM), which features the characteristic of loss in magnetism when the magnet temperature reaches the Curie point. A control rod with SASS is inserted into the core by gravity without recourse to any active shutdown system. To allow the SASS to effectively function, efficiently guiding high-temperature fluid from the fuel assembly to CPEM is important. Therefore, CPEM features a complicated shape such as having 45 fins, and a flow collector is installed upstream of CPEM to direct the flow from the fuel subassembly outlet to CPEM. In this report, the water experiment was performed on the full-scale model that simulates from the outlets of the fuel assemblies to the SASS flow collector, and flow phenomena around the temperature sensing part was analyzed from the data obtained by PIV measurement.

State sensing of bubble jet flow based on acoustic recognition and deep learning

Mikami, Nao*; Ueki, Yoshitaka*; Shibahara, Masahiko*; Aizawa, Kosuke; Ara, Kuniaki

International Journal of Multiphase Flow, 159, p.104340_1 - 104340_8, 2023/02

https://doi.org/10.1016/j.ijmultiphaseflow.2022.104340

This study covers the accidental generation of bubble jet flow caused by steam generator (SG) tubes damaging in sodium cooled fast reactors (SFRs). The main objective of this study is to develop a novel state sensing method of bubble jet flow based on acoustic recognition and deep learning. Prior to the application of this method to actual SFRs, we utilize air and water as simulant fluids in order to perform the proof of concept. This study is divided into three phases. The first phase is the acquisition and analysis of pipe flow sound and bubble jet flow sound, each of which simulates the normal and anomaly sound from SG tubes in SFRs. The second phase is the preprocessing of acoustic signals and feature extraction. The third phase is the building of deep learning models and performance evaluation. As a result, every of our proposed models could distinguish between pipe flow sound and bubble jet sound with an accuracy of almost 100.00%, and the best model could classify pipe flow sound and three types of bubble jet flow sound with an accuracy of 99.76%. This result suggests that the acoustic recognition with deep learning has great potential to sense the state of bubble jet flow in actual SFRs.

Investigation on natural circulation behavior for decay heat removal in reactor vessel of sodium-cooled fast reactor under severe accident condition, 2; Transient behavior under operations of multiple decay heat removal systems

Aizawa, Kosuke; Tsuji, Mitsuyo; Kobayashi, Jun; Kurihara, Akikazu

Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 7 Pages, 2022/10

In sodium-cooled fast reactors (SFRs), optimizing the design and operate decay heat removal systems (DHRSs) is important for safety enhancement against severe accidents. Thus, it is required to evaluate the cooling capability of DHRSs including the natural circulation behavior inside the reactor vessel during heat-removal phase that the fuel debris relocated in the reactor vessel is cooled by DHRSs. In this study, the experiments which simultaneously operations of the dipped-type DHX and the penetrated-type DHX were conducted to investigate the effect of operating multiple decay heat removal system on the natural circulation behavior in the reactor vessel. After achieving the stable conditions by operating the dipped-type DHX or the penetrated-type DHX, the other DHX was operated and the transient behavior was clarified by the temperature measurements. The clear temperature rise in the reactor vessel was confirmed by operating the penetrated-type DHX as second DHX operation under the condition of the dipped-type DHX operation at the beginning and the high heater power of fuel debris on the core catcher. Therefore, it was confirmed that the inhibition of the cooling for the decay heat occurred by operating multiple DHXs.

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

https://shop.elsevier.com/books/sodium-cooled-fast-reactors/morishita/978-0-12-824076-2

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.

Experimental study on velocity distribution in the subchannels of a fuel pin bundle with wrapping wire; Evaluation of the characteristics of flow field in 3-pin bundle

Hiyama, Tomoyuki; Aizawa, Kosuke; Nishimura, Masahiro; Kurihara, Akikazu

JAEA-Research 2021-009, 29 Pages, 2021/11

JAEA-Research-2021-009.pdf:2.25MB

In sodium-cooled fast reactors, high burnup of fuel is required for practical use. It is important to predict and evaluate the flow behavior in a fuel assembly because there is a concern that the heat removal capacity of the fuel assembly with high burnup will be locally reduced due to swirling and thermal deformation of the fuel rods. In this study, flow field measurement tests were conducted using a 3-pin bundle system test specimen for the purpose of elucidating the phenomenon and constructing a verification database for thermal hydraulics analysis code. The viewpoints of the experiment for elucidating the phenomenon are as follows; (1) Overall flow behavior in the subchannel including near the wrapping wire, (2) Relationship between Reynolds number including laminar flow region and flow field, and (3) Evaluation of the effect of the presence or absence of wrapping wire on the flow field. As a result, detailed flow field data in the subchannel was obtained by PIV measurement. It was found that when the wrapping wire crossed the subchannel, the flow occurred toward adjacent subchannel and the flow occurred that follows the winding direction of the wrapping wire. It was confirmed that the tendency of the flow velocity distribution of the Reynolds number in the laminar flow region is significantly different from that of the transition region and the turbulent region under the condition. The test was conducted using a same 3-pin bundle system without the wrapping wire, and it was confirmed that mixing by the wrapping wire occurred even in the laminar flow region.

Water experiments on thermal striping phenomena at the core outlet of an advanced sodium-cooled fast reactor, 2; Proposal of countermeasures to mitigate temperature fluctuations around radial blanket fuel assemblies

Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki

Hozengaku, 20(3), p.97 - 101, 2021/10

Focusing on the thermal striping phenomena that occurs at a bottom of the internal structure of an advanced sodium-cooled fast reactor (Advanced-SFR) that has been designed by the Japan Atomic Energy Agency, a water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of Upper Internal Structure (UIS). In the previous paper, we reported the effect of measures to mitigate temperature fluctuations around the control rod channels. In this paper, the same test section was used, and a water experiment was conducted to obtain the characteristics of temperature fluctuations around the radial blanket fuel assembly. And the shape of the Core Instrumentation Support Plate (CIP) was modified, and it was confirmed that it was highly effective in alleviating temperature fluctuations around the radial blanket fuel assembly.

Development of hydrogen monitor; Applicability investigation of acoustic technique

Aizawa, Kosuke; Ara, Kuniaki; Hino, Ryutaro; Hirabayashi, Masaru*

Proceedings of OECD/NEA Specialist Workshop on Advanced Measurement Method and Instrumentation for enhancing Severe Accident Management in an NPP addressing Emergency, Stabilization and Long-term Recovery Phases (SAMMI 2020) (Internet), 4 Pages, 2020/12

Many functions of the instrumentation system did not work in the Tokyo Electric Power Company Fukushima Daiichi Nuclear Power Plant in the severe condition. The function of the hydrogen concentration measurement system was lost due to power supply loss and coolant loss because the system was based on the sampling method. Therefore, the development of an on-site installation type hydrogen monitor which is not based on the sampling method and has environment resistance characteristics is required. Thus, a new type of hydrogen monitor by using acoustic technique has been developed at JAEA. The measurement principle is to detect the changing of sound velocity with the hydrogen concentration in a mixed gas. In this paper, the basic performance and influence of environmental conditions are described.

Investigation on velocity distribution in the subchannels of pin bundle with wrapping wire; Evaluation of Reynolds number dependence in 3-pin bundle

Aizawa, Kosuke; Hiyama, Tomoyuki; Nishimura, Masahiro; Kurihara, Akikazu; Ishida, Katsuji*

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 8 Pages, 2020/08

https://doi.org/10.1115/ICONE2020-16679

A sodium-cooled fast reactor is designed to attain a high burn-up core in commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, the deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity in the subassembly and influence the heat removal capability. Therefore, it is important to obtain the flow velocity distribution in a wire wrapped pin bundle. In this study, the detailed flow velocity distribution in the subchannel has been obtained by PIV (Particle Image Velocimetry) measurement using a wire-wrapped 3-pin bundle water model. Flow velocity conditions in the pin bundle were set from 0.036 m/s ( = 270) to 1.6m/s ( = 13,500). From the PIV results, the maximum flow velocity was increased by decreasing the number in the region away from the wrapping wire. Moreover, the PIV measurements by using the 3-pin bundle geometry without the wrapping wire were conducted. From the results, the effect of the wrapping wire on the flow field in the subchannel was understood. There experimental results useful not only for understanding of pin bundle thermal hydraulics but also code validation.