Numerical Simulation Research on Natural Circulation Flow of the Reactor Coolant System
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摘要: 采用计算流体动力学（CFD）程序建立了包含反应堆、蒸汽发生器、主泵和主管道在内的三环路反应堆冷却剂系统的高保真三维数值模型，开展了低功率运行工况下系统级热工水力现象的三维数值分析，获得了不同区域的冷却剂温度，并与核电厂实测数据对比，验证了数值模型的合理性。分析结果表明：该功率水平下的自然循环流量为满功率运行流量的4.5%，堆芯出口温度稳定，可以有效导出堆芯热量；局部热对流现象使不同环路的冷却剂产生更充分搅混；顶盖腔室内存在热分层现象，现有的顶盖温度测点读数不是该区域内的最高温度；主泵出口产生旋转流，并且靠近主管道管壁区域切向速度较大，中心区域形成局部对流。该研究工作可以进一步提升设计者对核电厂复杂系统级三维热工水力现象的认识。
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关键词:
- 计算流体动力学（CFD） /
- 反应堆冷却剂系统 /
- 自然循环
Abstract: Computational Fluid Dynamics(CFD) program is employed to enable the high-fidelity modeling of the reactor coolant system (RCS) for a typical three-loop pressurized water reactor, and the completed model of the RCS is build including reactor vessel and internals, core, steam generator, primary pump and linking pipe. The three-dimensional, global and localized flow features have been investigated under natural circulation flow condition with lower core thermal power, and the temperature at different locations are compared with the measured values from the operating nuclear power plant in order to verify the accurate description of the developed CFD model. The results show that the natural circulation flow rate is about 4.5% of the full power flow rate while the temperature of the core outlet is stable, and the residual core heat could be effectively removed. The phenomenon of the thermal stratification in the reactor pressure vessel head dome shows that the measured temperature value of the detector position in nuclear power plant could not provide the highest value. The coolant from different loops could be more fully mixed due to the local convection flow. There is a swirling flow at the outlet of primary pump, and the tangential velocity near the pipe wall is large while the local convection occurs at the central area. This analysis practice provides an effective evaluation for the system-level three-dimensional thermal hydraulics phenomena of the reactor coolant system. -
图 3 不同网格数量下的回路流量和顶盖旁流份额变化
Figure 3. Comparison of Loop Flow and Head bypass Percentage for Different Mesh
表 1 不同区域温度的计算值和实测值对比
Table 1. Temperature Comparison between CFD Values and Measured Values
参数名 冷管段
温度/℃热管段
温度/℃反应堆压力容器顶盖腔室
温度/℃堆芯出口
温度/℃实测值 294.8 295.6 294 296 计算值 294.7 296 294.8 296.1 相对偏差/% −0.03 0.14 0.27 0.03 
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表 2 不同区域温度的计算值和实测值对比
Table 2. Temperature Comparison between CFD Values and Measured Values
参数名 冷管段
温度/℃热管段
温度/℃反应堆压力容器顶盖腔室
温度/℃堆芯出口
温度/℃实测值 293 317.5 312 317.5 计算值 296.2 316.7 305.5 317.4 相对偏差/% 1.1 −0.25 −2.08 −0.03
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[1] International Atomic Energy Agency. Summary review on the application of computational fluid dynamics in nuclear power plant design: IAEA No. NR-T-1.20[R]. Vienna: International Atomic Energy Agency, 2022. [2] WANG M J, WANG Y J, TIAN W X, et al. Recent progress of CFD applications in PWR thermal hydraulics study and future directions[J]. Annals of Nuclear Energy, 2021, 150: 107836. doi: 10.1016/j.anucene.2020.107836 [3] 邢继,王辉,吴宇翔,等. 压水堆技术后续发展的思考[J]. 哈尔滨工程大学学报,2021, 42(12): 1707-1713. [4] 张明乾,冉小兵,刘言午,等. CPR1000反应堆三维数值模拟分析及验证[J]. 核技术,2013, 36(10): 100601. [5] 张明乾,段远刚,朱明莉,等. 基于CFD方法的反应堆流量分配结构的优化设计[J]. 核科学与工程,2015, 35(2): 193-199. doi: 10.3969/j.issn.0258-0918.2015.02.001 [6] PRASSER H M, KLIEM S. Coolant mixing experiments in the upper plenum of the ROCOM test facility[J]. Nuclear Engineering and Design, 2014, 276: 30-42. doi: 10.1016/j.nucengdes.2014.05.016 [7] KÜTÜK B, GÜZELBEY İ H. Computational fluid dynamics analyses of a VVER-1200 nuclear reactor vessel for symmetric inlet, asymmetric inlet, and LOCA conditions[J]. International Journal of Pressure Vessels and Piping, 2020, 187: 104165. doi: 10.1016/j.ijpvp.2020.104165 [8] BÖTTCHER M, BERNARD O, MAS A, et al. CFD analysis of coolant mixing in VVER-1000/V320 reactor pressure vessel[J]. Annals of Nuclear Energy, 2024, 197: 110274. doi: 10.1016/j.anucene.2023.110274 [9] ZHAO X H, WANG M J, CHEN C, et al. Three-dimensional study on the hydraulic characteristics under the steam generator (SG) tube plugging operations for AP1000[J]. Progress in Nuclear Energy, 2019, 112: 63-74. doi: 10.1016/j.pnucene.2018.10.016 [10] ZHANG X, WANG P F, RUAN X D, et al. Analysis of pressure pulsation induced by rotor-stator interaction in nuclear reactor coolant pump[J]. Shock and Vibration, 2017, 2017: 7363627. [11] 张恒,刘雷,刘立军. 非均匀入流对CAP1400核主泵内流及性能的影响研究[J]. 西安交通大学学报,2023, 57(2): 39-48. [12] 李瑶,刘建阁. 反应堆冷却剂系统主管道流场分析[J]. 核科学与技术,2018, 6(2): 20-26. [13] ZHANG W Q, WU C S, YANG S, et al. Study on the characteristics of the nuclear reactor coolant pump during the process-divided start-up period[J]. Annals of Nuclear Energy, 2022, 178: 109381. doi: 10.1016/j.anucene.2022.109381 [14] 黎义斌,张帆,郭艳磊,等. 反应堆一回路对核主泵叶轮入流特性的影响[J]. 排灌机械工程学报,2023, 41(10): 973-980. [15] MARTINEZ P, GALPIN J. CFD modeling of the EPR primary circuit[J]. Nuclear Engineering and Design, 2014, 278: 529-541. doi: 10.1016/j.nucengdes.2014.08.013 [16] NALBANDYAN A, CAMMI A, LORENZI S, et al. Modelling and CFD analysis of the DYNASTY loop facility[J]. EPJ- Nuclear Sciences & Technologies, 2022, 8: 12. [17] 王强,高璞珍,王忠乙,等. RELAP5对低压自然循环系统的分析能力研究[J]. 哈尔滨工程大学学报,2019, 40(5): 920-925. [18] 李楠,李浩永,张恒语,等. 基于RELAP5程序的单相自然循环数值计算敏感性分析[C]//第十五届全国反应堆热工流体学术会议暨中核核反应堆热工水力技术重点实验室学术年会论文集. 荣成: 中国核学会,2017. -
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