TY - JOUR
T1 - Void-fraction measurement with high spatial resolution in a 5×5 rod bundle by linear-accelerator-driven X-ray computed tomography over a wide pressure range
AU - Arai, Takahiro
AU - Furuya, Masahiro
AU - Takiguchi, Hiroki
AU - Nishi, Yoshihisa
AU - Shirakawa, Kenetsu
N1 - Funding Information:
Some of this research was conducted as part of the Infrastructure Development Project to Enhance Safety Measures at Nuclear Power Plants (Advanced Models for Thermal-hydraulic Analysis during the Nuclear Fuel Boil-off Process), which was sponsored by the Ministry of Economy, Trade and Industry in Japan. The authors would like to thank Messrs. Takio Endo, Yuki Miyazawa, and Yoshiyuki Shiratori of Electric Power Engineering Systems Co., Ltd. for their help with these experiments.
Funding Information:
Some of this research was conducted as part of the Infrastructure Development Project to Enhance Safety Measures at Nuclear Power Plants (Advanced Models for Thermal-hydraulic Analysis during the Nuclear Fuel Boil-off Process), which was sponsored by the Ministry of Economy, Trade and Industry in Japan. The authors would like to thank Messrs. Takio Endo, Yuki Miyazawa, and Yoshiyuki Shiratori of Electric Power Engineering Systems Co. Ltd. for their help with these experiments.
Publisher Copyright:
© 2019 Elsevier Ltd
PY - 2019/10
Y1 - 2019/10
N2 - Void fraction (i.e., the volume fraction occupied by gas) is a key parameter for determining the coolability and neutron-moderating performance of a water-cooled nuclear reactor. To develop computational multi-fluid dynamics models for determining the void-fraction distribution, experimental data of comparable quality are required. We have developed a high-energy X-ray computed tomography (CT) system to acquire three-dimensional void-fraction distributions. The CT system comprises a linear-accelerator-driven high-energy X-ray source and a linear detector array. We quantified a boiling two-phase flow in a 5 × 5 heated rod bundle at high pressure, simulating a fuel-rod bundle in a boiling water reactor (BWR). Because the axial travel of the CT system is 4 m and includes the entire BWR fuel-rod bundle, we optimized the CT imaging conditions and reconstruction method for rod-bundle visualization to reduce uncertainties due to density fluctuations in the boiling flow and imaging artifacts. We conducted a boiling experiment at a low flow rate and low thermal power and acquired three-dimensional distributions of the void fraction over a wide pressure range of 0.1–7.2 MPa. The experiment provided three-dimensional void-fraction distributions with high spatial resolution, especially in subchannel regions surrounded by rods, and the results are suitable for validating three-dimensional thermal-hydraulic analysis codes.
AB - Void fraction (i.e., the volume fraction occupied by gas) is a key parameter for determining the coolability and neutron-moderating performance of a water-cooled nuclear reactor. To develop computational multi-fluid dynamics models for determining the void-fraction distribution, experimental data of comparable quality are required. We have developed a high-energy X-ray computed tomography (CT) system to acquire three-dimensional void-fraction distributions. The CT system comprises a linear-accelerator-driven high-energy X-ray source and a linear detector array. We quantified a boiling two-phase flow in a 5 × 5 heated rod bundle at high pressure, simulating a fuel-rod bundle in a boiling water reactor (BWR). Because the axial travel of the CT system is 4 m and includes the entire BWR fuel-rod bundle, we optimized the CT imaging conditions and reconstruction method for rod-bundle visualization to reduce uncertainties due to density fluctuations in the boiling flow and imaging artifacts. We conducted a boiling experiment at a low flow rate and low thermal power and acquired three-dimensional distributions of the void fraction over a wide pressure range of 0.1–7.2 MPa. The experiment provided three-dimensional void-fraction distributions with high spatial resolution, especially in subchannel regions surrounded by rods, and the results are suitable for validating three-dimensional thermal-hydraulic analysis codes.
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U2 - 10.1016/j.flowmeasinst.2019.101614
DO - 10.1016/j.flowmeasinst.2019.101614
M3 - Article
AN - SCOPUS:85071717777
SN - 0955-5986
VL - 69
JO - Flow Measurement and Instrumentation
JF - Flow Measurement and Instrumentation
M1 - 101614
ER -