Study on Dynamical Stress Characteristics of DVI Nozzle Subjected to Pressurized Thermal Shocks

Zhao Yanyi; Wang Zewu; Li Junbao

doi:10.13832/j.jnpe.2022.04.0078

Volume 43 Issue 4

Aug. 2022

Turn off MathJax

Article Contents

Article Navigation > Nuclear Power Engineering > 2022 > 43(4): 78-85

Zhao Yanyi, Wang Zewu, Li Junbao. Study on Dynamical Stress Characteristics of DVI Nozzle Subjected to Pressurized Thermal Shocks[J]. Nuclear Power Engineering, 2022, 43(4): 78-85. doi: 10.13832/j.jnpe.2022.04.0078

Citation:

Zhao Yanyi, Wang Zewu, Li Junbao. Study on Dynamical Stress Characteristics of DVI Nozzle Subjected to Pressurized Thermal Shocks[J]. Nuclear Power Engineering, 2022, 43(4): 78-85. doi: 10.13832/j.jnpe.2022.04.0078

Citation:

PDF( 19668 KB)

Study on Dynamical Stress Characteristics of DVI Nozzle Subjected to Pressurized Thermal Shocks

doi: 10.13832/j.jnpe.2022.04.0078

Department of Chemical Machinery and Safety, Dalian University of Technology, Dalian, Liaoning, 116024, China

Received Date: 2021-08-02
Accepted Date: 2021-08-19
Rev Recd Date: 2021-08-18
Publish Date: 2022-08-04

Abstract

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.
- Reactor pressure vessel,
- DVI nozzle,
- Thermal-fluid-solid coupling,
- Pressurized thermal shocks,
- Structural integrity

FullText(HTML)

References(13)

References

[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