Study on Transient Hydraulic Load of Reactor Coolant System under the Condition of Reactor Coolant Pump Rotor Seizure
-
摘要: 为真实反映事故工况下反应堆冷却剂系统瞬态内流过渡过程及水力载荷冲击，针对“华龙一号”反应堆及一回路系统，建立了高精度三维闭式系统瞬态流动计算方法，得到了该过渡过程中反应堆及一回路系统管路内压力波震荡规律及瞬态水力载荷特性。研究表明：在核主泵转子卡滞结束时，核主泵出口处流量降低到稳定运行时的81.3%。在卡转子过渡过程中，系统管路内最大压力峰值位于核主泵入口截面处，为16.00 MPa，小压力谷值位于核主泵出口截面处，为15.01 MPa，最终系统内各监测点的压力趋于参考压力15.50 MPa。受反应堆冷却剂系统管路布置方式与核主泵卡转子事故的双重影响，各截面的流体速度呈现明显的不均匀分布，并产生了明显的涡流。系统各壁面水力载荷变化规律与系统压力脉动变化规律一致，最大载荷力峰值位于过渡段第一个弯头处的W3壁面处，为3.163×106 N；最小载荷力谷值位于反应堆压力容器入口处弯头的W12壁面处，为9.125×105 N。本数值预测方法可为主泵卡转子事故工况下反应堆冷却剂系统管路的设计与安全性评估提供技术支持。Abstract: In order to truly reflect the transient internal flow transition process and hydraulic load impact of the reactor coolant system under accident conditions, a high-precision three-dimensional closed system transient flow calculation method was established for the HPR1000 reactor and its primary system, and the pressure wave oscillation law and transient hydraulic load characteristics of the pipeline of the reactor and primary system during the transition process were obtained. The results show that: in the end of the reactor coolant pump rotor seizure, the flow rate at the reactor coolant pump outlet decreased to 81.3% of that in stable operation. During the transition process of the rotor seizure, the maximum pressure peak value in the system pipeline is located at the inlet section of the reactor coolant pump, which is 16.00 MPa; the minimum pressure valley value is located at the outlet section of the reactor coolant pump, which is 15.01 MPa. Finally, the pressure of each monitoring point in the system tends to the reference pressure of 15.50 MPa. Under the dual influence of the piping layout of reactor coolant system and the rotor seizure accident of reactor coolant pump, the fluid velocity of each section shows obvious uneven distribution, and obvious eddy currents occurs. The variation rule of hydraulic load on each wall of the system is consistent with the variation rule of system pressure pulsation. The maximum load force peak is located at the W3 wall at the first elbow of the transition section, which is 3.163×106 N; the minimum load force valley value is located at the W12 wall of the elbow at the inlet of the reactor pressure vessel, which is 9.125×105 N. This numerical prediction method can provide technical support for the design and safety assessment of the reactor coolant pipeline under the condition of the reactor coolant pump rotor seizure.
-
图 1 反应堆冷却剂系统三维简化示意图
Figure 1. Three Dimensional Simplified Schematic Diagram of Reactor Coolant System
图 4 反应堆冷却剂系统管路监测点及壁面示意图
Figure 4. Schematic Diagram of Reactor Coolant System Pipeline Monitoring Points and Wall Surfaces
图 8 反应堆一回路管路监测点的压力变化
Figure 8. Pressure Variation at Monitoring Points of Reactor Primary System Pipeline
图 10 核主泵进口不同截面的速度和流线分布图
Figure 10. Velocity and Streamline Distribution for Different Sections of Reactor Coolant Pump Inlet
图 11 核主泵出口不同截面的速度和流线分布图
Figure 11. Velocity and Streamline Distribution for Different Sections of Reactor Coolant Pump Outlet
-
[1] 刘昌文,李庆,李兰,等. “华龙一号”反应堆及一回路系统研发与设计[J]. 中国核电,2017, 10(4): 472-477,512. [2] 荆春宁,赵科,张力友,等. “华龙一号”的设计理念与总体技术特征[J]. 中国核电,2017, 10(4): 463-467. [3] 王永强,杨滨,武焕春,等. 揭秘核电材料——核电站一回路主管道材料及其制备工艺[J]. 金属世界,2013(1): 37-41. [4] 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 [5] 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 [6] LU Y G, ZHU R S, WANG X L, et al. Experimental study on transient performance in the coasting transition process of shutdown for reactor coolant pump[J]. Nuclear Engineering and Design, 2019, 346: 192-199. doi: 10.1016/j.nucengdes.2019.03.007 [7] 靖剑平,乔雪冬,贾斌,等. 基于RELAP5程序的AP1000典型事故瞬态特性研究[J]. 原子能科学技术,2015, 49(4): 646-653. [8] 齐炳雪,俞冀阳. 超临界水冷堆的安全分析[J]. 原子能科学技术,2012, 46(6): 669-673. [9] 王亚云. 核主泵及其系统过渡过程瞬态特性研究[D]. 大连: 大连理工大学,2018. [10] DIEN L D, DIEP D N. Verification of VVER-1200 NPP simulator in normal operation and reactor coolant pump coast-down transient[J]. World Journal of Engineering and Technology, 2017, 5(3): 507-519. doi: 10.4236/wjet.2017.53043 [11] 余红星,黄代顺. 秦山核电二期工程设计基准事故水力学载荷分析[J]. 核动力工程,2003, 24(S1): 102-105. [12] 李文姬,吕红,张洁. LOCA水力载荷分析软件HLPS的开发与验证[J]. 核动力工程,2021, 42(4): 159-165. [13] 卢喜丰,熊夫睿,刘文进,等. “华龙一号”蒸汽发生器传热管失水事故应力响应分析[J]. 压力容器,2021, 38(12): 70-76. [14] 肖文宇,范正伟. 小破口失水事故瞬态水力载荷计算中破口尺寸的影响[J]. 管道技术与设备,2020(6): 21-23. [15] 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. -
下载:
计量
- 文章访问数: 159
- HTML全文浏览量: 63
- PDF下载量: 29
- 被引次数: 0
