@article{cdd167ada7994d83a953f05f4bbdc33c,
title = "Development of Subchannel Void Sensor for Wide Pressure and Temperature Ranges and Its Application to Boiling Flow Dynamics in a Heated Rod Bundle",
abstract = "A subchannel void sensor (SCVS) acquires the two-phase flow in a rod bundle as the time-series data of cross-sectional distributions. Herein, the temperature and pressure ranges of an SCVS were extended to include the rated conditions of boiling water reactors. The improved SCVSs were installed in a 5 × 5 heated rod bundle at eight height levels. In a boiling experiment using the rod bundle, the three-dimensional distributions of the boiling two-phase flow were measured over a wide pressure range (up to 7.2 MPa). The new experimental data were compared with existing experimental data and the results of a subchannel analysis. Experimental results were consistent with those of a high-energy X-ray computed tomography study of a heated rod bundle with the same geometry and under the same heat and flow conditions as those used in our study. The subchannel analysis code reproduced the experimental results fairly well, and the obtained database is applicable for validating and improving thermal-hydraulic analysis codes.",
keywords = "Subchannel void sensor, boiling flow, rod bundle, subchannel analysis code, void fraction",
author = "Takahiro Arai and Masahiro Furuya and Kenetsu Shirakawa",
note = "Funding Information: Subchannel analysis codes are three-dimensional thermal-hydraulic analysis codes in a reactor core. The cross section of flow channel is divided into several tens to 200 subchannels depending on the rod bundle geometry. Herein, the experimental results were reproduced by the subchannel analysis code CTF, which is applicable to boiling two-phase flows in both BWRs and pressurized water reactors and includes the constitutive equations of the three separate fields (continuous liquid, vapor, and droplets in the vapor). CTF was developed by North Carolina State University under the Consortium for Advanced Simulation of Light-Water Reactors project (funded by the U.S. Department of Energy). CTF analyses of the void fraction distribution have been validated against the BFBT benchmark and the results of other experiments using rod bundles. However, because the sensor specifications and installation positions were limited at the time of the BFBT experiments, the X-ray CT measurements were conducted only at the top end of the heater rod. Conversely, the SCVS can acquire the two-phase flow parameters in the central subchannel and rod gap regions. The three-dimensional two-phase flows obtained in the experiment conducted herein are appropriate for validating the CTF. Funding Information: Part of this research was conducted as the Infrastructure Development Project for Enhancement of Safety Measures at Nuclear Power Plants “Advanced Models for Thermalhydraulic Analysis During Nuclear Fuel Boil-off Process” sponsored by the Ministry of Economy, Trade and industry. The authors would like to thank Takio Endo, Yuki Miyazawa, and Yoshiyuki Shiratori of the Electric Power Engineering Systems Co., Ltd. for their help in conducting these experiments. Funding Information: Part of this research was conducted as the Infrastructure Development Project for Enhancement of Safety Measures at Nuclear Power Plants ?Advanced Models for Thermalhydraulic Analysis During Nuclear Fuel Boil-off Process? sponsored by the Ministry of Economy, Trade and industry. The authors would like to thank Takio Endo, Yuki Miyazawa, and Yoshiyuki Shiratori of the Electric Power Engineering Systems Co., Ltd. for their help in conducting these experiments. Publisher Copyright: {\textcopyright} 2021 American Nuclear Society.",
year = "2022",
doi = "10.1080/00295450.2021.1897733",
language = "English",
volume = "208",
pages = "203--221",
journal = "Nuclear Technology",
issn = "0029-5450",
publisher = "American Nuclear Society",
number = "2",
}