Study on Hydrodynamic Characteristics of Transient Process of Reactor Coolant Pump Shaft Stuck Accident
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摘要: 为探究核主泵卡轴事故瞬变过程的水动力特性，通过动态匹配核主泵水力特性与系统管路阻力特性，建立了反应堆一回路系统的全三维简化模型。借助计算流体动力学（CFD）方法对核主泵卡轴事故工况进行了瞬态数值模拟，得到不同卡轴工况下核主泵外特性、内部压力场、叶轮叶片载荷与受力特性的瞬时变化。研究表明：卡轴时间越短，核主泵相应特性参数的瞬时变化越剧烈，事故造成影响越严重。以叶轮转速刚降为0 r/min时为节点，在卡轴时间为0.1、0.3、0.5 s三种卡轴工况下，流量分别降低到正常运行时的82.3%、61.4%、49.6%；核主泵扬程达到反向极值，分别为正常运行时的−137.7%、−87.4%、−56.9%；叶轮叶片两侧压力差值达到最大，分别为1.34、0.73、0.47 MPa，且在叶轮叶片工作面一侧和导叶流道中间部分形成相对集中的低压区；叶轮所受轴向力达到反向极值，分别为正常运行时的−159.3%、−96.5%、−65.5%。本数值预测方法对反应堆水动力系统的动态安全性评估提供了一定的数据支撑。Abstract: In order to explore the hydrodynamic characteristics of transient process of the reactor coolant pump shaft stuck accident, a full three-dimensional simplified model of the reactor primary circuit system was established by dynamically matching the hydraulic characteristics of the reactor coolant pump and the resistance characteristics of the system pipeline. The transient numerical simulation of the reactor coolant pump shaft stuck accident condition is carried out by using the computational fluid dynamics (CFD) method, and the transient variations of external characteristics, internal pressure field, impeller blade load and force of the reactor coolant pump under different shaft stuck conditions are obtained. The study shows that the shorter the shaft stuck time, the more dramatic the transient variation of the reactor coolant pump characteristic parameters, and the more serious the impact of the accident. Taking the moment when the impeller speed just drops to 0 r/min as the node, under three shaft stuck conditions (i.e. shaft stuck time = 0.1 s, 0.3 s and 0.5 s), the flow rate decreases to 82.3%, 61.4% and 49.6%, respectively, of that under the normal operation. The head of the reactor coolant pump reaches the reverse extreme value, i.e. −137.7%, −87.4% and −56.9% , respectively, of the value under the normal operation. The pressure difference between the two sides of the impeller blade reaches the maximum, i.e. 1.34 MPa, 0.73 MPa and 0.47 MPa, respectively, and a relatively concentrated low-pressure area is formed on the side of the blade working surface in the impeller blade and in the middle part of the guide vane flow channel. The reverse extreme value of the axial force on the impeller reaches −159.3%, −96.5% and −65.5%, respectively, of the value under the normal operation. The numerical prediction method provides certain data support for the dynamic safety assessment of the reactor hydrodynamic system.
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图 2 反应堆一回路系统简化示意图
Figure 2. Simplified Schematic Diagram of Reactor Primary Circuit System
图 7 不同卡轴工况下系统流量与转速的变化
Figure 7. Variation of System Flow and Speed under Different Shaft Stuck Conditions
图 8 不同卡轴工况下核主泵扬程与转速的变化
Figure 8. Variation of Reactor Coolant Pump Head and Speed under Different Shaft Stuck Conditions
图 9 叶轮及导叶中间流线切面压力云图
Figure 9. Pressure Nephogram in Middle Streamline Section of Impeller and Guide Vane
图 10 叶轮叶片中间流线切面压力分布
Figure 10. Pressure Distribution in Middle Streamline Section of Impeller Blade
图 11 不同卡轴工况下叶轮径向力变化
Figure 11. Variation of Impeller Radial Force under Different Shaft Stuck Conditions
图 12 不同卡轴工况下叶轮轴向力的变化
Figure 12. Variation of Impeller Axial Force under Different Shaft Stuck Conditions
表 1 核主泵额定参数
Table 1. Rated Parameters of Reactor Coolant Pump
参数名 参数值 额定流量/(m3·h−1) 24680 设计扬程/m 102.8 额定转速/(r·min−1) 1485 参考压力/MPa 15.5 叶轮叶片数 4 导叶叶片数 11 
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[1] 邓华,王秋颖,伍鹏,等. 核电站一回路主系统三维仿真模型的研究与应用[J]. 工业控制计算机,2016, 29(1): 77-79. doi: 10.3969/j.issn.1001-182X.2016.01.034 [2] 龙云,袁寿其,朱荣生,等. 核主泵内部流动研究现状与技术发展综述[J]. 排灌机械工程学报,2020, 38(11): 1081-1097. [3] 刘永. 小破口卡轴事故工况下AP1000核主泵水动力特性分析[D]. 镇江: 江苏大学, 2017. [4] 钟伟源. 卡轴事故工况下核主泵流固耦合瞬变特性研究[D]. 镇江: 江苏大学, 2018. [5] LI D H, LI Y B, ZHANG F, et al. Study on hydraulic characteristics of reactor coolant pump shutdown transition process based on primary circuit closed system[J]. Frontiers in Energy Research, 2022, 9: 808393. doi: 10.3389/fenrg.2021.808393 [6] TANG J X, LU D G, LIANG J T, et al. Numerical simulation on asymmetrical three-dimensional thermal and hydraulic characteristics of the primary sodium pool under the pump stuck accident in CEFR[J]. Nuclear Science and Engineering, 2021, 195(5): 478-495. doi: 10.1080/00295639.2020.1834314 [7] WANG X L, LU Y G, ZHU R S, et al. Study on the transient evolution law of internal flow field and dynamic stress of reactor coolant pump under rotor seizure accident[J]. Annals of Nuclear Energy, 2019, 133: 35-45. doi: 10.1016/j.anucene.2019.05.001 [8] AZZOUNE M, BOUMEDIEN A, LABABSA D, et al. Analysis of a loss-of-flow accident resulting from the primary pump shaft break transient of the NUR research reactor[J]. Journal of Nuclear Science and Technology, 2019, 56(1): 130-145. doi: 10.1080/00223131.2018.1532845 [9] 苏宋洲,王鹏飞,许忠斌,等. 核主泵启动过程压力脉动和径向力研究[J]. 核动力工程,2017, 38(3): 110-114. doi: 10.13832/j.jnpe.2017.03.0110 [10] 王秀礼,袁寿其,朱荣生,等. 核主泵停机过渡过程瞬态水动力特性研究[J]. 原子能科学技术,2013, 47(3): 364-370. doi: 10.7538/yzk.2013.47.03.0364 [11] 付强,袁寿其,朱荣生,等. 失水事故工况下核主泵气液两相瞬态流动特性[J]. 华中科技大学学报:自然科学版,2013, 41(9): 112-116. [12] GAO H, GAO F, ZHAO X C, et al. Analysis of reactor coolant pump transient performance in primary coolant system during start-up period[J]. Annals of Nuclear Energy, 2013, 54: 202-208. doi: 10.1016/j.anucene.2012.11.020 [13] YE D X, LAI X D, LUO Y M, et al. Diagnostics of nuclear reactor coolant pump in transition process on performance and vortex dynamics under station blackout accident[J]. Nuclear Engineering and Technology, 2020, 52(10): 2183-2195. doi: 10.1016/j.net.2020.03.015 [14] 辛素芳,李松蔚,任春明,等. 蒸汽发生器U型管CFD简化方法研究[J]. 核动力工程,2018, 39(S1): 41-44. doi: 10.13832/j.jnpe.2018.S1.0041 [15] 何戈宁,张富源,李冬慧,等. 华龙一号ZH-65型蒸汽发生器设计研发与性能验证[J]. 压力容器,2021, 38(4): 48-52. doi: 10.3969/j.issn.1001-4837.2021.04.007 [16] 汪建,杨帅,吴大转,等. 基于三维CFD方法的管路瞬变流特性研究[J]. 流体机械,2019, 47(4): 6-13. doi: 10.3969/j.issn.1005-0329.2019.04.002 -
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