压水堆管道小破口蒸汽临界流泄漏实验研究

朱梦馨; 殷松涛; 王海军; 王宁宁

doi:10.13832/j.jnpe.2023.02.0084

压水堆管道小破口蒸汽临界流泄漏实验研究

doi: 10.13832/j.jnpe.2023.02.0084

1.
西安交通大学动力工程多相流国家重点实验室，西安，710049
2.
比亚迪股份有限公司，广东深圳，518118

详细信息

作者简介:
朱梦馨（1998—），女，硕士研究生，现主要从事热工水力研究，Email: zhumengxin@stu.xjtu.edu.cn

通讯作者:
王海军，E-mail: whj@mail.xjtu.edu.cn

中图分类号: TL334
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- 被引次数: 0
出版历程
- 收稿日期: 2022-04-21
- 修回日期: 2022-06-28
- 刊出日期: 2023-04-15

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

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

摘要

摘要: 为探究压水堆核电厂小破口失水事故中管道小破口蒸汽临界流泄漏特性，开展了管道小破口泄漏实验，以探索饱和/过热蒸汽临界流泄漏特性。基于压力管道疲劳贯穿裂纹（微通道），开展了流体压力3~12 MPa、流体温度240℃~320℃范围内的蒸汽临界流泄漏实验。实验结果表明，蒸汽临界质量流速与初始流体压力呈正相关关系，与初始流体过热度呈负相关关系。与过冷水临界流泄漏相比，蒸汽临界质量流速受入口压力损失、摩擦效应与加速效应的影响相对较弱。利用一维等熵模型预测了蒸汽临界质量流速，预测值与实验值平均相对偏差为14.17%，表明一维等熵模型具有良好的蒸汽临界质量流速预测精度。
- 失水事故 /
- 蒸汽泄漏 /
- 两相临界流 /
- 疲劳裂纹
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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图 1 实验装置图

Figure 1. Experimental Apparatus Diagram

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图 2 实验段示意图

Figure 2. Schematic Diagram of Experimental Section

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图 3 实验件示意图

Figure 3. Schematic Diagram of Test Piece

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图 4 压力对临界质量流量的影响

Figure 4. Influence of Pressure on Critical Mass Flow

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图 5 压力对临界质量流速的影响

Figure 5. Influence of Pressure on Critical Mass Flow Rate

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图 6 过热度对临界质量流速的影响

Figure 6. Influence of Superheat on Critical Mass Flow Rate

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图 7 压力对干度的影响

Figure 7. Influence of Pressure on Mass Quality

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图 8 模型预测值与实验值比较

Figure 8. Comparison of the Predicted Value of the Model and the Experimental Value

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表 1 实验件编号及尺寸

Table 1. Numbers and Dimensions of Test Pieces

参数	实验件尺寸
参数	编号1	编号2	编号3
管道壁厚/mm	10	10	10
裂纹深度/mm	7.2	7.2	7.2
外壁面裂纹长度/mm	17.25	13.90	10.37
内壁面裂纹长度/mm	20	20	20
内壁面COD/mm	0.12	0.16	0.10
外壁面COD/mm	0.120	0.110	0.061
出口面积/mm²	2.07	1.53	0.63

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表 2 关键参数不确定度

Table 2. Uncertainty of Key Parameters

参数	不确定度/%
温度	0.24~0.25
压力	0.74
流量	0.92

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参考文献(14)

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