Study on Dynamical Stress Characteristics of DVI Nozzle Subjected to Pressurized Thermal Shocks
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摘要: 压力容器直接注入（DVI）接管在热冲击下的动态应力特性对于反应堆压力容器（RPV）结构完整性评估具有重要意义。建立了含DVI接管的RPV压力壳热流固耦合数值计算模型，并进行了验证分析；然后研究了蓄压安注箱（ACC）和堆芯补水箱（CMT）安注时RPV筒体和DVI接管热工水力特性；最后分析了热冲击下RPV筒体和DVI接管连接高应力区的温度分布、等效应力和等效塑性应变分布特性。研究结果表明，ACC安注阶段RPV筒体和DVI接管连接区存在较大的温度梯度和等效应力，且发生了局部塑性变形。若发生承压热冲击事件，应控制好DVI接管连接区温差，确保反应堆压力容器的结构完整性。本文开发的热冲击下热流固耦合数值计算模型和计算方法可用于核岛内DVI接管与RPV筒体的安全性评价，也可用于类似承压结构在热冲击下的动态应力特性分析。Abstract: The dynamic stress characteristics of direct vessel injection (DVI) nozzle under thermal shock are of great significance for the structural integrity evaluation of reactor pressure vessel (RPV). The thermal-fluid-solid coupling numerical calculation model of the RPV pressurized shell containing the DVI nozzle is firstly established and verified. Then the thermal hydraulic characteristics of RPV cylinder and DVI nozzle of ACC and CMT are studied; Finally, the distribution characteristics of temperature, equivalent stress and equivalent plastic strain in the high stress zone of RPV cylinder and DVI nozzle under thermal shock are analyzed. The results show that there is a large temperature gradient and equivalent stress in the connection area between RPV cylinder and DVI nozzle during ACC safety injection, and local plastic deformation occurs. In case of pressurized thermal shock event, the temperature difference in DVI nozzle area shall be controlled to ensure the structural integrity of reactor pressure vessel. The thermal-fluid-solid coupling numerical calculation model and method developed in this paper can be used for the safety evaluation of DVI nozzle and RPV cylinder in nuclear island, and can also be used to analyze the dynamic stress characteristics of similar pressure-bearing structure under thermal shock.
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图 5 流固耦合传热模型示意图
$ {T_{\text{s}}} $—固体网格中心点温度；$ {T_{\text{f}}} $—流体网格中心点温度；$ {T_{\text{w}}} $—壁面第一层网格中心点温度；$ {{\Delta }}{y_{\text{f}}} $—流体网格中心点到壁面距离；${{\Delta }}{y_{\text{s}}}$—固体网格中心点到壁面距离
Figure 5. Schematic Diagram of Fluid-solid Coupling Heat Transfer Model
图 6 RPV压力壳流固耦合传热模型
Figure 6. Fluid-solid Coupling Heat Transfer Model of RPV Pressurized Shell
图 7 YOZ平面流体温度及速度分布
Figure 7. Temperature and Velocity Distributions of the Fluid at the YOZ Plane
图 8 冷却过程中RPV压力壳温度分布云图
Figure 8. Temperature Contour of the RPV Pressurized Shell during the Cooling Process
图 9 冷却过程中关键位置温度变化曲线
Figure 9. Temperature Variation Curves of Key Points during the Cooling Process
图 11 t=200 s时等效塑性应变及局部放大图
Figure 11. Equivalent Plastic Strain and Local Enlarged View at t=200 s
图 12 冷却过程中最大等效塑性应变曲线
Figure 12. Curve of the Maximum Equivalent Plastic Strain during Cooling Process
表 1 SA508Gr.3材料参数
Table 1. Material Parameters of SA508Gr.3
温度/
℃弹性
模量/
GPa切线
模量/
GPa屈服
强度/
MPa热膨胀
系数/
10−6℃−1比热容/
(J·kg−1·℃−1)热导率/
(W·m−1·℃−1)0 193 9.8 437 11.3 400 41.4 100 187 9.8 413 12.1 483 40.6 200 181 9.8 394.3 12.8 526 40 300 174 9.8 381 13.2 566 38.8 
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表 2 安注水及冷却剂流量和温度参数
Table 2. Flow and Temperature Parameters of the Safety Injection Water and Coolant
事件 时间（t）/s 流量
/(kg·s−1)温度/℃ ACC安注 0<t≤100 690 10 CMT再注入 100<t<200 78 246 冷却剂注入 0<t<200 63.3 263
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表 3 监测点A和A*处h值的理论与数值计算结果比较
Table 3. Comparison between Theoretical and Numerical Solutions of h Values of the Points A and A*
监测点
位置DVI接管入口流
量/(kg·s−1)h计算结果/(W·m−2·k−1) 相对误
差/%理论 数值 A 690 49672.06 45246.04 9.78 78 7173.07 6565.11 9.26 A* 690 49672.06 46818.25 6.10 78 7173.07 6681.74 7.35
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[1] HUANG P C, CHOU H W, FERNG Y M, et al. Large thermal gradients on structural integrity of a reactor pressure vessel subjected to pressurized thermal shocks[J]. International Journal of Pressure Vessels and Piping, 2020, 179: 103942. doi: 10.1016/j.ijpvp.2019.103942 [2] 左巧林,秋穗正,王明军. 恰希玛核电厂压力容器各关键部位在PTS瞬态下的温度场研究[J]. 原子能科学技术,2016, 50(5): 829-834. [3] 张丽屏,苏东川,高世卿,等. 反应堆压力容器接管嘴内隅角应力强度因子计算研究[J]. 原子能科学技术,2017, 51(11): 2042-2048. doi: 10.7538/yzk.2017.youxian.0275 [4] 蒋兴,贺寅彪,张明. 承压热冲击瞬态下反应堆压力容器下降环腔内三维热工水力分析[J]. 压力容器,2020, 37(4): 46-49. doi: 10.3969/j.issn.1001-4837.2020.04.007 [5] 牛莉莎,叶红光,施惠基. 承压热冲击对核压力容器强度的影响[J]. 核动力工程,2001, 22(3): 242-247. doi: 10.3969/j.issn.0258-0926.2001.03.011 [6] 刘开泰, 黄美, 程建. 承压热冲击对AP1000反应堆压力容器的影响[C]//北京力学会第二十五届学术年会会议论文集. 北京: 北京力学会, 2019: 2. [7] SHARABI M, GONZÁLEZ-ALBUIXECH V F, LAFFERTY N, et al. Computational fluid dynamics study of pressurized thermal shock phenomena in the reactor pressure vessel[J]. Nuclear Engineering and Design, 2016, 299: 136-145. doi: 10.1016/j.nucengdes.2015.10.014 [8] UITSLAG-DOOLAARD H J, STEFANINI L, SHAMS A, et al. Numerical prediction of a single phase pressurized thermal shock scenario for crack assessment in an reactor pressure vessel wall[J]. Annals of Nuclear Energy, 2020, 144: 107563. doi: 10.1016/j.anucene.2020.107563 [9] GONZÁLEZ-ALBUIXECH V F, QIAN G, SHARABI M, et al. Comparison of PTS analyses of RPVs based on 3D-CFD and RELAP5[J]. Nuclear Engineering and Design, 2015, 291: 168-178. doi: 10.1016/j.nucengdes.2015.05.025 [10] 秦勉,于涛,于德勇,等. 承压热冲击下AP1000压力容器直接安注瞬态数值模拟研究[J]. 核动力工程,2015, 36(1): 1-8. [11] 马梓淇,张宏亮,何培峰,等. 瞬态承压热冲击下AP1000反应堆压力容器热应力数值模拟[J]. 中国科学:技术科学,2017, 47(7): 685-691. [12] 王宝生,王冬青,董化平,等. 全厂断电事故下AP1000非能动余热排出系统瞬态特性数值分析[J]. 原子能科学技术,2013, 47(9): 1514-1521. doi: 10.7538/yzk.2013.47.09.1514 [13] 吴海玲,罗毓珊,李悦,等. 承压热冲击下压水堆压力容器壁面换热特性的数值模拟研究[J]. 核动力工程,2002, 23(2): 27-33. doi: 10.3969/j.issn.0258-0926.2002.02.006 -
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