Evaluation of the Impact of Core Power Distribution on Reflooding Phenomenon in CCTF Test
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摘要: 以圆柱形堆芯试验装置（CCTF）为研究对象，采用轻水堆冷却系统事故工况的瞬态行为最佳估算程序（RELAP5）和自主化堆工设计与安全分析程序（LOCUST），开展堆芯功率分布对CCTF C2-SH2（Run54）试验工况再淹没现象影响的评价研究。研究表明：①计算所得下降段压降、堆芯压降、堆芯出口蒸汽质量流量等计算结果与试验结果吻合较好；②对于堆芯1.015 m处平均通道包壳峰值温度的计算，RELAP5和LOCUST程序计算的包壳峰值温度分别为816 K和813 K，试验结果为898 K，计算值比试验值低约82 K，平均通道包壳温度最后稳定在400 K左右，计算结果与试验结果一致。因此，本研究结果表明LOCUST程序能够较好地对大破口失水事故（LBLOCA）中再淹没阶段的瞬态过程进行模拟。
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关键词:
- 圆柱形堆芯试验装置（CCTF） /
- RELAP5 /
- LOCUST /
- 再淹没 /
- 堆芯功率
Abstract: The cylindrical core test facility (CCTF) is taken as the research object, optimal estimation program for transient behavior of LWR cooling system under accident condition RELAP5 and autonomous core design and safety analysis program LOCUST are used to evaluate the impact of core power distribution on the reflooding phenomenon under CCTF C2-SH2 (Run54) test conditions. The research shows: ① The calculations, such as the pressure drop in the descending section and core and the steam mass flowrate at the outlet of the code, provide good agreement with the test results; ② For the calculation of the average channel cladding peak temperature at 1.015 m of the core, the cladding peak temperatures calculated by RELAP5 and LOCUST programs are 816 K and 813 K respectively, the test results are 898 K, the calculated value is about 82 K lower than the test value, and the average channel cladding temperature finally stabilized at about 400 K. The calculated results are consistent with the test results. Therefore, the results of this study show that the LOCUST program can better simulate the transient process of the reflooding phase in the large-break loss-of-coolant accident (LBLOCA).-
Key words:
- CCTF /
- RELAP5 /
- LOCUST /
- Reflooding /
- Core power
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图 3 下降段整体压降变化曲线
Figure 3. Variation Curve of Overall Pressure Drop in Descending Section
图 6 堆芯平均通道1.015 m处包壳温度变化曲线
Figure 6. Variation Curve of Cladding Temperature at 1.015 m of Core Average Channel
表 1 CCTF C2-SH2（Run54）试验参数表[3]
Table 1. Test Parameters Table of CCTF C2-SH2(Run54)
序号 参数 数值 1 堆芯功率/MW 7.87 2 安全壳压力/MPa 0.2 3 蒸汽发生器二次侧压力/MPa 5.3 4 ECC冷却剂温度/K 310 5 蒸汽发生器二次侧温度/K 539 6 蓄压安注注水流量/(m3·s−1) 注入下腔室 0.104 注入冷管段 0.088 7 低压安注注水流量/(m3·s−1) 0.011 
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表 2 CCTF压力容器建模类型和数量
Table 2. CCTF Pressure Vessel Modeling Type and Quantity
区域 部件类型 数量 流体区域 环管部件（Annulus） 8 分支部件（Branch） 10 圆管部件（Pipe） 12 多连接件（Multiple Junction） 6 单一控制体（Single Volume） 2 单一连接件（Single Junction） 2 时间控制体（Time Dependent Volume） 1 时间连接件（Time Dependent Junction） 1 固体区域 热构件（Heat Structures） 44
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表 3 CCTF C2-SH2（Run54）工况初始条件和边界条件参数表[3]
Table 3. Initial Condition and Boundary Condition Parameter Table of CCTF C2-SH2 (Run54) Working Condition
序号 参数 数值 1 总功率/MW 7.87 2 平均线功率/(kW·m−1) 1.18 3 堆芯轴向功率分布（A∶B∶C） 1.37∶1.20∶0.76 4 安全壳压力/MPa 0.2 5 蒸汽发生器模拟器二次侧压力/MPa 5.3 6 下降段壁面温度/K 468 7 主环路管道壁面温度/K 406 8 蒸汽发生器模拟器二次侧温度/K 539 9 下腔室冷却剂温度/K 394 10 ECC冷却剂温度/K 310 11 压力容器内部温度/K 423 12 下腔室水位/m 0.86 13 蒸汽发生器模拟器二次侧水位/m 7.4 14 蓄压安注注水速度/(m3·s−1) 注入下腔室 0.104 注入冷段 0.088 15 低压安注注水速度/(m3·s−1) 0.011
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[1] OKUBO T, IGUCHI T, MURAO Y. Experimental study on difference in reflood core heat transfer among CCTF, FLECHT-SET and predicted with FLECHT correlation[J]. Journal of Nuclear Science and Technology, 1994, 31(8): 839-849. doi: 10.1080/18811248.1994.9735230 [2] IWAMURA T, OSAKABE M, SUDO Y. Effects of radial core power profile on core thermo-hydraulic behavior during reflood phase in PWR-LOCAs[J]. Journal of Nuclear Science and Technology, 1983, 20(9): 743-751. doi: 10.1080/18811248.1983.9733461 [3] IGUCHI T, SUGIMOTO J, AKIMOTO H, et al. Evaluation report on CCTF Core-II reflood test second shakedown test C2-SH2 (Run54): JAERI-M 85-025[R]. Japan: Japan Atomic Energy Research Institute, 1985. [4] IGUCHI T, MURAO Y, AKIMOTO H. Assessment of core radial power profile effect model for REFLA code by using CCTF data[J]. Journal of Nuclear Science and Technology, 1987, 24(7): 536-546. doi: 10.1080/18811248.1987.9735844 -
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