Experimental Study on Steam Critical Flow Leakage from a Small Break in Pipeline of Pressurized Water Reactor

Zhu Mengxin; Yin Songtao; Wang Haijun; Wang Ningning

doi:10.13832/j.jnpe.2023.02.0084

Volume 44 Issue 2

Apr. 2023

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Zhu Mengxin, Yin Songtao, Wang Haijun, Wang Ningning. Experimental Study on Steam Critical Flow Leakage from a Small Break in Pipeline of Pressurized Water Reactor[J]. Nuclear Power Engineering, 2023, 44(2): 84-90. doi: 10.13832/j.jnpe.2023.02.0084

Citation:

Zhu Mengxin, Yin Songtao, Wang Haijun, Wang Ningning. Experimental Study on Steam Critical Flow Leakage from a Small Break in Pipeline of Pressurized Water Reactor[J]. Nuclear Power Engineering, 2023, 44(2): 84-90. doi: 10.13832/j.jnpe.2023.02.0084

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Experimental Study on Steam Critical Flow Leakage from a Small Break in Pipeline of Pressurized Water Reactor

doi: 10.13832/j.jnpe.2023.02.0084

1.
State Key Laboratory of Multiphase Flow in Power Engineering, Xi'an Jiaotong University, Xi'an, 710049, China
2.
BYD Company Limited, Shenzhen, Guangdong, 518118, China

Received Date: 2022-04-21
Rev Recd Date: 2022-06-28
Publish Date: 2023-04-15

Abstract

Abstract

In order to explore the characteristics of steam critical flow leakage from a small break in pipeline during the loss of coolant accident of pressurized water reactor (PWR) nuclear power plant, small-break leakage experiments of pipelines are carried out in this paper to explore the characteristics of saturated/superheated steam critical flow leakage. Based on pressure pipeline fatigue through crack (microchannel), the steam critical flow leakage experiment is carried out within the fluid pressure range of 3~12 MPa and the fluid temperature range of 240℃~320℃. The experimental results show that the critical mass flow rate of steam is positively correlated with the initial fluid pressure and negatively correlated with the initial fluid superheat degree. Compared with the critical flow leakage of supercooled water, the critical mass flow rate of steam is less affected by inlet pressure loss, friction effect and acceleration effect. The one-dimensional isentropic model is used to predict the critical mass flow rate of steam. The mean relative deviation between the predicted value and the experimental value is 14.17%, which indicates that the one-dimensional isentropic model can accurately predict the critical mass flow rate of steam.
- Loss of coolant accident,
- Steam leakage,
- Two-phase critical flow,
- Fatigue crack

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References(14)

References

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