Study on the Drift Flux Model of Two-Phase Flow in the Liquid Metal Pool

Mou Zhuoya; Zhu Longxiang; Wang Di; Ouyang Yong; Zhang Hong; Wan Lingfeng; Zhang Luteng; Tang Simiao; Pan Liangming

doi:10.13832/j.jnpe.2024.09.0024

Volume 46 Issue 5

Oct. 2025

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Article Navigation > Nuclear Power Engineering > 2025 > 46(5): 92-100

Mou Zhuoya, Zhu Longxiang, Wang Di, Ouyang Yong, Zhang Hong, Wan Lingfeng, Zhang Luteng, Tang Simiao, Pan Liangming. Study on the Drift Flux Model of Two-Phase Flow in the Liquid Metal Pool[J]. Nuclear Power Engineering, 2025, 46(5): 92-100. doi: 10.13832/j.jnpe.2024.09.0024

Citation:

Mou Zhuoya, Zhu Longxiang, Wang Di, Ouyang Yong, Zhang Hong, Wan Lingfeng, Zhang Luteng, Tang Simiao, Pan Liangming. Study on the Drift Flux Model of Two-Phase Flow in the Liquid Metal Pool[J]. Nuclear Power Engineering, 2025, 46(5): 92-100. doi: 10.13832/j.jnpe.2024.09.0024

Citation:

Mou Zhuoya, Zhu Longxiang, Wang Di, Ouyang Yong, Zhang Hong, Wan Lingfeng, Zhang Luteng, Tang Simiao, Pan Liangming. Study on the Drift Flux Model of Two-Phase Flow in the Liquid Metal Pool[J]. Nuclear Power Engineering, 2025, 46(5): 92-100. doi: 10.13832/j.jnpe.2024.09.0024

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Study on the Drift Flux Model of Two-Phase Flow in the Liquid Metal Pool

doi: 10.13832/j.jnpe.2024.09.0024

Mou Zhuoya^{1, 2
,},
Zhu Longxiang^{1, 2
,
,},
Wang Di³,
Ouyang Yong³,
Zhang Hong^{1, 2},
Wan Lingfeng^{1, 2},
Zhang Luteng^{1, 2},
Tang Simiao^{1, 2},
Pan Liangming^{1, 2}

1.
Key Laboratory of Low-Grade Energy Utilization Technologies and Systems, Ministry of Education, Chongqing University, Chongqing, 400044, China
2.
Department of Nuclear Engineering and Technology, Chongqing University, Chongqing, 400044, China
3.
China Nuclear Power Technology Research Institute Co., Ltd., Shenzhen, Guangdong, 518000, China

Received Date: 2024-09-29
Accepted Date: 2024-11-13
Rev Recd Date: 2024-11-13

Available Online: 2025-10-15

Publish Date: 2025-10-15

Abstract

Abstract

When the Steam Generator Tube Rupture (SGTR) occurs in lead-cooled fast reactors, the gas-liquid metal two-phase flow in the lead pool of primary circuit poses a severe threat to the safe operation of reactors. Understanding the flow characteristics of bubbles in the liquid metal pool and developing a drift flux model applicable to liquid metal two-phase flow are of great significance for improving the accuracy of reactor core safety predictions. An experimental study on the local distribution characteristics of phase parameters in a Woods alloy liquid pool was conducted using the double-sensor conductivity probe measurement technique. The experimental results show that the radial distribution of phase parameters such as void fraction tends to develop towards a core-peaked profile as the superficial gas velocity increases. Based on two-phase flow experimental data, a comparative analysis was conducted to evaluate the applicability of existing gas-water two-phase flow drift flux models in predicting the void fraction of liquid metal two-phase flow. It was found that only a few models showed good agreement with the experimental void fraction data. Combined with the experimental data and comparison results, the key parameters in the distribution parameter and drift velocity were modified based on the existing model. The relative calculation error of the newly developed model is within ±15%, and the absolute relative error is 5.35%.
- Lead-cooled fast reactors,
- Steam Generator Tube Rupture (SGTR),
- Gas-liquid metal two-phase flow,
- Drift flux model,
- Void fraction,
- Conductivity probe

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References(30)

References

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