Mechanical Analysis of Pressure Vessels and Main Pipes under Steam Explosion Loads
-
摘要: 针对华龙一号反应堆压力容器（RPV），研究其在假设蒸汽爆炸载荷下RPV和主管的力学响应。通过建立有限元模型并根据瞬态结构分析方法开展数值分析，得到了RPV和主管道的变形、应力和应变结果。计算结果表明：RPV在600、800、1000℃下的失效载荷分别为1/20、1/50和1/100设计载荷；最大等效应力/应变均位于接管附近；主管道大部分区域应力未超过管道屈服应力。本研究可为RPV极端载荷下的结构完整性分析提供技术支持。
-
关键词:
- 反应堆压力容器（RPV） /
- 蒸汽爆炸 /
- 主管道 /
- 数值分析
Abstract: The mechanical response of HPR1000 reactor pressure vessel (RPV) and its main pipe under the hypothetical steam explosion load is studied. Through the establishment of the finite element model and the numerical analysis according to the transient structure analysis method, the deformation, stress and strain results of RPV and the main pipeline are obtained. The results show that the failure loads of RPV at 600℃, 800℃ and 1000℃ are 1/20, 1/50 and 1/100 design loads respectively; The maximum equivalent stress/strain is located near the nozzle; The stress in most areas of the main pipe does not exceed the yield stress of the pipe. This study can provide technical support for structural integrity analysis of RPV under extreme load.-
Key words:
- Reactor pressure vessel (RPV) /
- Steam explosion /
- Main pipe /
- Numerical analysis
-
图 2 不同位置的蒸汽爆炸设计压力载荷
Figure 2. Design Pressure Load for Steam Explosion of Different Locations
图 3 不同载荷下RPV等效应力-时间曲线
Figure 3. Equivalent Stress-Time Curves of RPV under Different Loads
图 9 不同载荷下支撑最大等效应力曲线
Figure 9. Maximum Equivalent Stress of Support under Different Loads
表 1 力学参数
Table 1. Mechanical Parameter
温度/℃ 弹性模量/GPa 屈服应力/MPa 抗拉强度/MPa 热导率/[W·(m·℃)−1] 比热容/[J·(kg·℃)−1] 热膨胀系数/10−6℃−1 20 200 475 594 42.02 0.517 10.8 100 190 416 544 44.10 0.537 12.4 200 190 398 542 44.10 0.577 14.0 300 190 423 593 42.02 0.616 14.9 400 170 369 512 40.47 0.676 16.4 600 118 243 276 36.32 0.875 14.4 800 30 50 74 24.64 0.736 18.1 1000 36 18 30 47.90 2.699 18.7 1050 30 14 25 47.90 2.699 20.6 
下载: 导出CSV
-
[1] U. S. Nuclear Regulatory Commission, Steam Explosion Review Group. A review of current understanding of the potential for containment failure arising from in-vessel steam explosion: NUREG-1116[R]. Washington: U. S. Nuclear Regulatory Commission, 1985. [2] THEOFANOUS T G. The study of steam explosions in nuclear systems[J]. Nuclear Engineering and Design, 1995, 155(1-2): 1-26. doi: 10.1016/0029-5493(94)00864-U [3] LESKOVAR M, URŠIČ M. Estimation of ex-vessel steam explosion pressure loads[J]. Nuclear Engineering and Design, 2009, 239(11): 2444-2458. doi: 10.1016/j.nucengdes.2009.07.023 [4] LESKOVAR M, CENTRIH V, URŠIČ M. Simulation of steam explosion in stratified melt-coolant configuration[J]. Nuclear Engineering and Design, 2016, 296: 19-29. doi: 10.1016/j.nucengdes.2015.10.026 [5] AHN K I, PARK S H, KIM H D, et al. The plant-specific uncertainty analysis for an ex-vessel steam explosion-induced pressure load using a TEXAS–SAUNA coupled system[J]. Nuclear Engineering and Design, 2012, 249: 400-412. doi: 10.1016/j.nucengdes.2012.04.015 [6] MORIYAMA K, PARK H S. Probability distribution of ex-vessel steam explosion loads considering influences of water level and trigger timing[J]. Nuclear Engineering and Design, 2015, 293: 292-303. doi: 10.1016/j.nucengdes.2015.07.062 [7] GRISHCHENKO D, GALUSHIN S, KUDINOV P. Failure domain analysis and uncertainty quantification using surrogate models for steam explosion in a Nordic type BWR[J]. Nuclear Engineering and Design, 2019, 343: 63-75. doi: 10.1016/j.nucengdes.2018.12.013 [8] 黄熙,杨燕华,王溪. 堆外蒸汽爆炸堆腔压力冲量分布计算分析[J]. 核动力工程,2011, 32(3): 15-21. [9] 钟明君,林萌,张政铭,等. 基于1000 MW级压水堆核电厂压力容器外蒸汽爆炸的模拟研究[J]. 核动力工程,2014, 35(4): 43-47. [10] 张荣金, 白伟, 倪伟峰, 等. 核电站堆外蒸汽爆炸影响因素探讨[C]//第十八届中国科协年会——分6军民融合高端论坛论文集. 西安: 中国科学技术协会学会学术部, 2016: 189-196. [11] 李春,杨志义,丁超,等. 基于MC3D软件对核电厂压力容器蒸汽爆炸的重要参数计算及研究[J]. 核安全,2018, 17(2): 58-65. -
计量
- 文章访问数: 94
- HTML全文浏览量: 50
- PDF下载量: 9
- 被引次数: 0
