Study on CCFL Characteristics in Downcomer during Discharge Phase of LOCA with RELAP5 Code
-
摘要: 反应堆失水事故（LOCA）后下降段通道内形成的两相逆流状态极有可能引发汽-液逆向流动限制（CCFL），不利于应急冷却水顺利进入堆芯，极大影响了核反应堆系统的安全性能。本研究基于RELAP5程序采用Wallis溢流关系式对UPFT实验装置进行建模并计算LOCA喷放阶段的下降段注水行为；通过对比下腔室蓄水量、下降段内压力及破口处蒸汽流量瞬态变化以验证模型的有效性，并对下降段通道内汽相速度场、液相体积分数分布特性进行分析。结果表明，由于下降段通道结构的三维特征引起的流动不均匀性影响了汽-液CCFL特性，随着蒸汽流量增大，在破口环路与下降段连接区域的压力梯度与向上流速度梯度越大，较少节点的划分方法很难真实反映下降段通道局部区域内汽-液溢流关系；在靠近破口的环路内注入的冷却水更难到达下腔室，而在远离破口环路的冷却水容易进入到下腔室；过热的蒸汽在流动过程中被冷却水冷却发生凝结现象，导致出口蒸汽流量小于进口蒸汽流量，且随着进口蒸汽流量的增大，凝结效应则随之减小。本研究所建立的模型与方法能够适用于LOCA喷放阶段下降段通道内的汽-液CCFL预测。
-
关键词:
- 应急冷却剂 /
- 下降段 /
- 逆向流动限制（CCFL） /
- 下腔室蓄水量 /
- RELAP5
Abstract: Under the loss-of-coolant accident (LOCA), the two-phase countercurrent in downcomer is extremely likely to cause the vapor-liquid counter-current flow limitation (CCFL), which is not conducive to the smooth entry of emergency coolant into the core, which greatly affects the safety performance of the nuclear reactor system. Based on RELAP5 code, the Wallis overflow relation is used to model the UPFT experimental device and calculate the water injection behavior in the downcomer during discharge phase of LOCA; The validity of the model is verified by comparing the water storage capacity of lower chamber, the pressure in the downcomer and the transient changes of steam flow at the break, and the distribution characteristics of vapor phase velocity field and liquid phase volume fraction in the downcomer are analyzed. The results show that the flow irregularity caused by the three-dimensional characteristics of the channel structure in the downcomer affects the characteristics of the vapor-liquid CCFL. With the increase of steam flow, the greater the pressure gradient and upward flow velocity gradient in the connection area between the break loop and the downcomer, the division method with fewer nodes is difficult to truly reflect the vapor-liquid overflow relationship in the local area of the downcomer channel; The cooling water injected in the circuit close to the break is more difficult to reach the lower chamber, while the cooling water in the circuit far from the break can easily enter the lower chamber; The superheated steam is cooled by the cooling water during the flow process, resulting in condensation, resulting in the steam flow at the outlet being less than the inlet steam flow, and the condensation effect decreases with the increase of the inlet steam flow. The model and method established in this study can be applied to the prediction of vapor-liquid CCFL in the downcomer channel during discharge phase of LOCA. -
图 1 UPTF Test 6工况下RELAP5模型节点图
Figure 1. Node Diagram of RELAP5 Model under Condition of UPTF Test 6
图 3 蒸汽喷放后5 s下降段内轴向汽流速度分布云图
Figure 3. Contours of Axial Steam Velocity Distribution in the Downcomer 5 s after Steam Discharge
图 4 下腔室蓄水量计算结果与实验值对比
Figure 4. Comparison between the Calculation Results of the Water Storage Capacity of Lower Chamber and the Experimental Values
图 5 冷却水注入15 s后下降段内液相体积分数云图
Figure 5. Contours of Liquid Phase Volume Fraction in the Downcomer after Cooling Water Injection for 15 s
图 6 下降段压力计算值与实验值对比
Figure 6. Comparison between Calculated Value and Experimental Value of Pressure in the Downcomer
图 7 破口处蒸汽流量计算值与实验值对比
Figure 7. Comparison between Calculated Value and Experimental Value of Steam Flow at the Break
表 1 UPTF Test 6工况参数
Table 1. UPTF Test 6 Condition Parameters
工况编号 单个蒸汽发生器注
汽量/(kg·s−1)堆芯模拟器注汽量/
(kg·s−1)CL1、CL3、CL4的ECCS注
水量/(kg·s−1)ECCS温度/K 安全壳模拟器
背压/kPa压力容器初始
压力/kPa131 29.0 309 482 391.15 252 244 132 30.0 205 490 385.15 248 250 133 31.0 110 491 390.15 256 257 136 0.0 102 490 387.15 240 245 
下载: 导出CSV
-
[1] AL ISSA S, MACIAN R. A review of CCFL phenomenon[J]. Annals of Nuclear Energy, 2011, 38(9): 1795-1819. doi: 10.1016/j.anucene.2011.04.021 [2] AL ISSA S, MACIAN-JUAN R. Experimental investigation and CFD validation of countercurrent flow limitation (CCFL) in a large-diameter PWR hot-leg geometry[J]. Journal of Nuclear Science and Technology, 2016, 53(5): 647-655. doi: 10.1080/00223131.2015.1125312 [3] GLAESER H, ROHATGI U S. Scaling ability of the counter-current flow limitation (CCFL) correlations for application to reactor thermal hydraulics[J]. Nuclear Engineering and Design, 2019, 354: 110226. doi: 10.1016/j.nucengdes.2019.110226 [4] CHEN C Y, SHIH C, WANG J R, et al. Sensitivity study on the counter-current flow limitation in the DEG LBLOCA with the TRACE code[J]. Annals of Nuclear Energy, 2013, 57: 121-129. doi: 10.1016/j.anucene.2013.01.025 [5] NIKITIN K, MUELLER P, MARTIN J, et al. BWR loss of coolant accident simulation by means of RELAP5[J]. Nuclear Engineering and Design, 2016, 309: 113-121. doi: 10.1016/j.nucengdes.2016.09.008 [6] TAKEDA T, OHTSU I. RELAP5 uncertainty evaluation using ROSA/LSTF test data on PWR 17% cold leg intermediate-break LOCA with single-failure ECCS[J]. Annals of Nuclear Energy, 2017, 109: 9-21. doi: 10.1016/j.anucene.2017.05.007 [7] 江灼威,蔡杰进. LOCA事故时一回路冷却剂管肘部回流流动极限研究[J]. 核科学与工程,2019, 39(3): 414-422. doi: 10.3969/j.issn.0258-0918.2019.03.011 [8] SIDDIQUI H, BANERJEE S, ARDRON K H. Flooding in an elbow between a vertical and a horizontal or near-horizontal pipe: Part I: Experiments[J]. International Journal of Multiphase Flow, 1986, 12(4): 531-541. doi: 10.1016/0301-9322(86)90058-3 [9] DAMERELL P S, SIMONS J W. 2D/3D program work summary report: NUREG/IA-0126[R]. Washington, DC: Nuclear Regulatory Commission, 1993 [10] DAMERELL P S, SIMONS J W. Reactor safety issues resolved by the 2D/3D Program: NUREG/IA-0127[R]. Washington, DC: Nuclear Regulatory Commission, 1993. [11] U. S. NRC. RELAP5/MOD3.3 code manual Vol. 1: code structure, system models, and solution methods. Nuclear safety analysis division Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission: RELAP5/MOD3.3 Code Manual: NUREG/CR-5535[R]. Rockville: Information Systems Laboratories, Inc. , 2001 [12] WANG M J, ZHAO H, ZHANG H P, et al. Research on the designed emergency passive residual heat removal system during the station blackout scenario for CPR1000[J]. Annals of Nuclear Energy, 2012, 45: 86-93. doi: 10.1016/j.anucene.2012.03.004 [13] WALLIS G B. One-dimensional two-phase flow[M]. New York: McGraw-Hill, 1969. -
计量
- 文章访问数: 216
- HTML全文浏览量: 255
- PDF下载量: 37
- 被引次数: 0
