Research on Crack Failure of Auxiliary Feed Water Pump Impeller Based on CFD

Li Yuanzheng; Liu Zheng; Zheng Hongen

Volume 41 Issue 2

Apr. 2020

Turn off MathJax

Article Contents

Abstract

References

Article Navigation > Nuclear Power Engineering > 2020 > 41(2): 130-134

Li Yuanzheng, Liu Zheng, Zheng Hongen. Research on Crack Failure of Auxiliary Feed Water Pump Impeller Based on CFD[J]. Nuclear Power Engineering, 2020, 41(2): 130-134.

Citation:

Li Yuanzheng, Liu Zheng, Zheng Hongen. Research on Crack Failure of Auxiliary Feed Water Pump Impeller Based on CFD[J]. Nuclear Power Engineering, 2020, 41(2): 130-134.

Li Yuanzheng, Liu Zheng, Zheng Hongen. Research on Crack Failure of Auxiliary Feed Water Pump Impeller Based on CFD[J]. Nuclear Power Engineering, 2020, 41(2): 130-134.

Citation:

Li Yuanzheng, Liu Zheng, Zheng Hongen. Research on Crack Failure of Auxiliary Feed Water Pump Impeller Based on CFD[J]. Nuclear Power Engineering, 2020, 41(2): 130-134.

PDF( 0 KB)

Research on Crack Failure of Auxiliary Feed Water Pump Impeller Based on CFD

Publish Date: 2020-04-11

Abstract

Abstract

In order to study the crack failure characteristics of the auxiliary feed pump impeller, the pressure pulsation response law of the auxiliary feed pump casing measuring point under different impeller crack lengths is explored based on CFD, and the CFD results are verified by the vibration monitoring data of the auxiliary pump feed pump in the presence or absence of impeller cracks. The research shows that the pressure pulsation response of the pump casing can better reflect the crack failure characteristics of the auxiliary feed pump impeller; when there is a crack in the auxiliary feed pump impeller, there will be a clear sideband with a width of rotating frequency of the blade passing frequency in the vibration spectrum of the pump casing measuring point.
- Auxiliary feed pump,
- Computational fluid dynamics (CFD),
- Impeller crack,
- Unsteady flow field simulation,
- Vibration monitoring

FullText(HTML)

References(0)

References

Relative Articles

[1]	Zhang Jing, Wang Mingjun, Tian Wenxi, Su Guanghui, Qiu Suizheng. Application of Artificial Intelligence Algorithms in Thermal-Hydraulic Analysis of Nuclear Reactors[J]. Nuclear Power Engineering, 2025, 46(2): 127-140. doi: 10.13832/j.jnpe.2024.090039
[2]	Chen Yongchao, Wei Xingfang, Liu Yanwu, Fang Jian, Ran Xiaobing. Experimental and CFD Simulation of the RPV Head Plenum[J]. Nuclear Power Engineering. doi: 10.13832/j.jnpe.2024.070004
[3]	Chen Xi, Wang Xiaoyu, Cui Cong, Deng Jian, Liu Yu, Liu Luguo, Liang Yu, Peng Huanhuan. Research on Thermal Diffusion Performance of Fuel Assembly Spacer Grid Based on Numerical Methodology[J]. Nuclear Power Engineering, 2024, 45(6): 91-97. doi: 10.13832/j.jnpe.2024.06.0091
[4]	Zhang Ji, Wang Yingjie, Wang Mingjun, Tian Wenxi, Qiu Suizheng, Su Guanghui. Study on Jet Mixing Characteristics of Lead-Bismuth Eutectic Cooled Reactor Assembly Head Based on CFD Method[J]. Nuclear Power Engineering, 2024, 45(4): 77-86. doi: 10.13832/j.jnpe.2024.04.0077
[5]	Deng Jian, Wei Zonglan, Zeng Wei, Li Songwei, Qiu Zhifang, Liu Luguo. Development and Application of Autonomous Computational Fluid Dynamics Code WINGS-CFD for Nuclear Reactors[J]. Nuclear Power Engineering, 2024, 45(S2): 63-69. doi: 10.13832/j.jnpe.2024.S2.0063
[6]	Zhao Hongwei, Dong Rui, Li Junjie, Ouyang Zhengrong, Heng Yuekun. Numerical Simulation of Displacement Process of Liquid Scintillator and Ultrapure Water in a Large Diameter Spherical Tank[J]. Nuclear Power Engineering, 2024, 45(2): 225-230. doi: 10.13832/j.jnpe.2024.02.0225
[7]	Liu Zhenglong, Qiu Hanrui, Wang Mingjun, Sun Hao, Tian Wenxi, Su Guanghui. Simulation of Blockage Accident of LBE-Cooled Fast Reactor Fuel Assembly Based on Mesh Deformation Method[J]. Nuclear Power Engineering, 2024, 45(3): 95-103. doi: 10.13832/j.jnpe.2024.03.0095
[8]	Xin Sufang, Wu Dan, Shen Yaou, Ren Chunming, Chen Shilong. Characterization Analysis of Core Outlet Temperature Measurement in Nuclear Power Plants[J]. Nuclear Power Engineering, 2023, 44(3): 237-242. doi: 10.13832/j.jnpe.2023.03.0237
[9]	Lu Dingsheng, Wang Chen, Wang Chenglong, Yue Nina, Yang Ping, Tian Wenxi, Su Guanghui, Qiu Suizheng. Study on Oxidation Corrosion Characteristics of Horizontal Lead-Bismuth Reactor Core[J]. Nuclear Power Engineering, 2023, 44(3): 96-103. doi: 10.13832/j.jnpe.2023.03.0096
[10]	Zeng Fulin, Zhang Xiaolong, Zhao Pengcheng. Study on Circumferential Non-Uniformity of Annular Fuel Outer Temperature Distribution in Lead-bismuth Cooled Tight Cell[J]. Nuclear Power Engineering, 2023, 44(6): 95-103. doi: 10.13832/j.jnpe.2023.06.0095
[11]	Wen Shuang, Zeng Xiehu, Wen Qinglong, Ruan Shenhui, Zhang Ruiqian, Wei Tianguo, Yang Hongyan. Numerical Study on Typical Cell of Fuel Assembly by Turbulence Excitation[J]. Nuclear Power Engineering, 2023, 44(1): 9-16. doi: 10.13832/j.jnpe.2023.01.0009
[12]	Su Qianhua, Fan Guanhua, Lyu Lulu, Lu Donghua, Yang Ping, Gan Fujun, Yan Binghuo, Wang Chengyue. Effect of the Arrangement for Grid Spacers with Mixing Vanes on the Thermal Hydraulic Characteristics of a 5×5 Rod Bundle via CFD Analysis[J]. Nuclear Power Engineering, 2023, 44(4): 88-94. doi: 10.13832/j.jnpe.2023.04.0088
[13]	Yang Tao, Zhao Pengcheng, Zhao Yanan, Yu Tao. Development of Reduced-Order Thermal Stratification Model for Upper Plenum of Lead-Bismuth Fast Reactor Based on CFD[J]. Nuclear Power Engineering, 2023, 44(2): 48-53. doi: 10.13832/j.jnpe.2023.02.0048
[14]	Li Jinyang, Ma Jun, Qiao Shouxu, Hao Sijia, Li Xupeng, Tan Sichao, Tian Ruifeng. Investigation on Characteristics of Resistance and Flow Distribution of 5×5 Annular Fuel Rod Bundle Channel under Pulsating Flow Condition[J]. Nuclear Power Engineering, 2023, 44(6): 111-118. doi: 10.13832/j.jnpe.2023.06.0111
[15]	Chen Mingrui, Wei Zonglan, Chen Chong, Deng Jian, Zhu Li, Peng Shinian. Numerical Simulation of Flow Heat Transfer in Rectangular Channel under Local Deformation Condition[J]. Nuclear Power Engineering, 2023, 44(3): 223-230. doi: 10.13832/j.jnpe.2023.03.0223
[16]	Liang Ruixian, Yang Lingfeng, Wang Yifeng, Li Xiaobo, Niu Fenglei. Numerical Study on Oxygen Transport Characteristics in Liquid Lead-Bismuth Eutectic Circuit[J]. Nuclear Power Engineering, 2022, 43(6): 187-194. doi: 10.13832/j.jnpe.2022.06.0187
[17]	Ren Bing, Gan Fujun, Yang Ping. CFD Calculation and Analysis of Boiling Heat Transfer of Full-length Fuel Bundle with Bent Rod[J]. Nuclear Power Engineering, 2022, 43(6): 43-50. doi: 10.13832/j.jnpe.2022.06.0043
[18]	Hu Shiqu, Gu Kai, Sun Xinyu, Yuan Jingqi, Dong He, Hui Jiuwu. Numerical Simulations of Flow Field in First-Stage Steam-water Separator of Steam Generator[J]. Nuclear Power Engineering, 2022, 43(5): 7-11. doi: 10.13832/j.jnpe.2022.05.0007
[19]	Hou Jianfei, Wang Qingli, Si Hengyuan. Analysis of Post-Accident Debris Transport Performance in In-Containment Refueling Water Storage Tank in Nuclear Power Plants[J]. Nuclear Power Engineering, 2021, 42(5): 76-80. doi: 10.13832/j.jnpe.2021.05.0076
[20]	Jiang Hong, Zhou Jing, Liu Lizhi. Numerical Analysis on the Characteristics of Steam Condensation in Presence of Air under Vertical Tube Bundle Conditions[J]. Nuclear Power Engineering, 2019, 40(5): 41-45.