铅铋堆蒸汽发生器传热管破裂事故三维程序开发及验证

辜峙钘; 余红星; 黄代顺; 严明宇; 申亚欧; 冯文培; 龚政宇

doi:10.13832/j.jnpe.2023.04.0226

铅铋堆蒸汽发生器传热管破裂事故三维程序开发及验证

doi: 10.13832/j.jnpe.2023.04.0226

1.
中国核动力研究设计院核反应堆系统设计技术重点实验室，成都，610213
2.
成都理工大学核技术与自动化工程学院，成都，610059

基金项目: 国家重点研发计划项目（2019YFB1901300）；国家自然科学基金项目（12005025）；四川省自然科学基金项目（2022NSFSC0253，2022NSFSC1233）

详细信息

作者简介:
辜峙钘（1987—），男，副教授，博士后（在站），现从事反应堆热工流体与安全分析相关研究，E-mail: guzhixing17@163.com

通讯作者:
余红星，E-mail: hongxing_yu@126.com

中图分类号: TL334
计量
- 文章访问数: 188
- HTML全文浏览量: 25
- PDF下载量: 36
- 被引次数: 0
出版历程
- 收稿日期: 2023-04-12
- 修回日期: 2023-05-30
- 刊出日期: 2023-08-15

Development and Verification of 3D Code for Steam Generator Tube Rupture Accident of LBE-cooled Reactor

1.
Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, 610213, China
2.
College of Nuclear Technology and Automation Engineering, Chengdu University of Technology, Chengdu, 610059, China

摘要

摘要: 蒸汽发生器（SG）传热管破裂事故（SGTR）是铅铋堆设计必须重点考虑的安全问题之一。针对铅铋堆SGTR，为解决其复杂结构环境中压力波的三维传播与蒸汽的三维迁移难题，基于多相流欧拉流体动力学理论，开展了“铅铋-水”相互作用三维数值模型与算法研究，研制了专用程序，并采用实验对比和程序对比技术手段进行了程序验证，验证结果符合较好。研究结果表明：对于描述铅铋堆SGTR过程中“铅铋-水”相互作用行为，本文采用的相关数值理论与模型具有较好的适用性；对于研究复杂结构环境下铅铋堆SGTR的三维演化现象，包括压力波传播、蒸汽迁移，本文所开发的三维程序具有重要的潜在应用价值。本文研究成果有望为我国铅铋堆SGTR分析提供有力支撑。
- 铅铋堆 /
- 蒸汽发生器（SG）传热管破裂事故（SGTR） /
- "铅铋-水"相互作用 /
- 三维程序开发 /
- 程序验证
Abstract: Steam Generator Tube Rupture (SGTR) accident is one of the significant safety problems that must be considered in the design of Lead-Bismuth-Eutectic (LBE) cooled reactor. With respect to the SGTR in LBE-cooled reactor, and to cope with the challenges of 3D propagation of pressure waves and 3D migration of water steam under the complex geometric structures, the three-dimensional numerical model and algorithm of the interaction between LBE and water were studied based on the Euler hydrodynamic theory of multiphase flow, and a special code was developed. The code was verified by means of experimental comparison and code-to-code comparison, and the verification results were in good agreement. It is demonstrated that the numerical theories and models used in this paper are suitable for describing the "LBE-water" interaction during SGTR accidents in LBE-cooled reactors. And the 3D code developed in this paper has important potential application values in coping with the 3D evolution processes of SGTR accidents in LBE-cooled reactors under the complex geometric structures, including the pressure wave propagation and steam migration. The research achievements in this paper are expected to provide strong supports for the SGTR accident analyses of LBE-cooled reactors in China.
- LBE-cooled reactor /
- SGTR /
- "LBE-water" interaction /
- 3D Code development /
- Code verification.

HTML全文

图 1 流型划分

白色连续填充为气体,蓝色与灰色连续填充为液体(铅铋或水),白色圈为气泡，蓝色与灰色圈为液滴(铅铋或水)

Figure 1. Flow Pattern Division

下载: 全尺寸图片幻灯片

图 2 “水注铅铋”实验装置单位：mm

Figure 2. Experimental Apparatus of Water Injection into LBE (mm)

下载: 全尺寸图片幻灯片

图 3 液体相对体积份额

y—笛卡尔三维坐标系中的y坐标

Figure 3. Liquid Components Relative Volume Fraction

下载: 全尺寸图片幻灯片

图 4 LIFUS5/MOD2实验台架

Figure 4. LIFUS5/MOD2 Experimental Facility

下载: 全尺寸图片幻灯片

图 5 压力演化过程

Figure 5. Variation Process of Pressure

下载: 全尺寸图片幻灯片

图 6 空气内压力波传播三维传播过程

Figure 6. 3D Process of Pressure Wave Propagation in Air

下载: 全尺寸图片幻灯片

图 7 单相水蒸汽内压力波传播过程

Figure 7. Process of Pressure Wave Propagation in Single-phase Water Steam

下载: 全尺寸图片幻灯片

图 8 激波管压力分布

Figure 8. Pressure Distribution in Shock Tube

下载: 全尺寸图片幻灯片

图 9 激波管速度分布

Figure 9. Velocity Distribution in Shock Tube

下载: 全尺寸图片幻灯片

图 10 激波管蒸汽体积份额分布

Figure 10. Volume Fraction Distribution in Shock Tube

下载: 全尺寸图片幻灯片

参考文献(25)

[1]	ALEMBERTI A, SMIRNOV V, SMITH C F, et al. Overview of lead-cooled fast reactor activities[J]. Progress in Nuclear Energy, 2014, 77: 300-307. doi: 10.1016/j.pnucene.2013.11.011
[2]	DIERCKS D R, SHACK W J, MUSCARA J. Overview of steam generator tube degradation and integrity issues[J]. Nuclear Engineering and Design, 1999, 194(1): 19-30. doi: 10.1016/S0029-5493(99)00167-3
[3]	CINOTTI L, SMITH C F, SEKIMOTO H, et al. Lead-cooled system design and challenges in the frame of Generation IV International Forum[J]. Journal of Nuclear Materials, 2011, 415(3): 245-253. doi: 10.1016/j.jnucmat.2011.04.042
[4]	WANG G. A review of research progress in heat exchanger tube rupture accident of heavy liquid metal cooled reactors[J]. Annals of Nuclear Energy, 2017, 109: 1-8. doi: 10.1016/j.anucene.2017.05.034
[5]	SMITH C F, CINOTTI L, SEKIMOTO H. Lead-cooled fast reactor (LFR) ongoing R&D and key issues: LLNL-CONF-414709[R]. Livermore, CA: Lawrence Livermore National Lab. , 2009.
[6]	DINH T N. Multiphase flow phenomena of steam generator tube rupture in a lead-cooled reactor system: a scoping analysis[C]//Proceedings of ICAPP. Nice. 2007.
[7]	SIBAMOTO Y, KUKITA Y, NAKAMURA H. Visualization and measurement of subcooled water jet injection into high-temperature melt by using high-frame-rate neutron radiography[J]. Nuclear Technology, 2002, 139(3): 205-220. doi: 10.13182/NT02-A3314
[8]	SIBAMOTO Y, KUKITA Y, NAKAMURA H. Small-scale experiment on subcooled water jet injection into molten alloy by using fluid temperature-phase coupled measurement and visualization[J]. Journal of Nuclear Science and Technology, 2007, 44(8): 1059-1069. doi: 10.1080/18811248.2007.9711347
[9]	CIAMPICHETTI A, AGOSTINI P, BENAMATI G, et al. LBE–water interaction in sub-critical reactors: First experimental and modelling results[J]. Journal of Nuclear Materials, 2008, 376(3): 418-423. doi: 10.1016/j.jnucmat.2008.02.039
[10]	PESETTI A, DEL NEVO A, FORGIONE N. Experimental investigation and SIMMER-III code modelling of LBE–water interaction in LIFUS5/Mod2 facility[J]. Nuclear Engineering and Design, 2015, 290: 119-126. doi: 10.1016/j.nucengdes.2014.11.016
[11]	CIAMPICHETTI A, BERNARDI D, CADIOU T, et al. LBE–water interaction in LIFUS 5 facility under different operating conditions[J]. Journal of Nuclear Materials, 2011, 415(3): 449-459. doi: 10.1016/j.jnucmat.2011.04.051
[12]	SA R, TAKAHASHI M, MORIYAMA K. Study on fragmentation behavior of liquid lead alloy droplet in water[J]. Progress in Nuclear Energy, 2011, 53(7): 895-901. doi: 10.1016/j.pnucene.2011.05.003
[13]	SA R, TAKAHASHI M. Experimental study on thermal interaction of ethanol jets in high temperature fluorinert[J]. Journal of Power and Energy Systems, 2012, 6(2): 314-323. doi: 10.1299/jpes.6.314
[14]	黄望哩. 铅基堆SGTR事故下铅铋与水接触碎化行为研究[D]. 合肥: 中国科学技术大学, 2015.
[15]	ZHANG C D, SA R Y, ZHOU D N, et al. Effects of gas velocity and break size on steam penetration depth using gas jet into water similarity experiments[J]. Progress in Nuclear Energy, 2017, 98: 38-44. doi: 10.1016/j.pnucene.2017.02.006
[16]	WANG S, FLAD M, MASCHEK W, et al. Evaluation of a steam generator tube rupture accident in an accelerator driven system with lead cooling[J]. Progress in Nuclear Energy, 2008, 50(2-3): 363-369.
[17]	GU Z X, GANG W, BAI Y Q, et al. Preliminary investigation on the primary heat exchanger lower head rupture accident of forced circulation LBE-cooled fast reactor[J]. Annals of Nuclear Energy, 2015, 81: 84-90. doi: 10.1016/j.anucene.2015.03.018
[18]	辜峙钘. 铅基研究实验堆无保护瞬态安全特性分析[D]. 合肥: 中国科学技术大学, 2017.
[19]	ZHANG C D, SA R Y, ZHOU D N, et al. Effects of failure location and pressure on the core voiding under SGTR accident in a LBE-cooled fast reactor[J]. International Journal of Heat and Mass Transfer, 2019, 141: 940-948. doi: 10.1016/j.ijheatmasstransfer.2019.06.106
[20]	陈家铭. 铅冷快堆中换热器泄漏瞬态事故下气泡迁移行为研究[D]. 合肥: 中国科学技术大学, 2020.
[21]	陈家铭,陈红丽. 自然循环铅冷快堆蒸汽发生器泄漏事故下的气泡迁移[J]. 原子能科学技术,2020, 54(12): 2344-2352. doi: 10.7538/yzk.2020.youxian.0004
[22]	HUANG X, CHEN B, YING Y, et al. Numerical investigation on LBE-water interaction for heavy liquid metal cooled fast reactors[J]. Nuclear Engineering and Design, 2020, 361: 110567. doi: 10.1016/j.nucengdes.2020.110567
[23]	YU Q F, ZHANG Y, WANG C L, et al. Numerical simulation of bubble transport during steam generator tube rupture accident of Lead-cooled Fast Reactor[J]. Annals of Nuclear Energy, 2021, 153: 108066. doi: 10.1016/j.anucene.2020.108066
[24]	于启帆,赵亚峰,王成龙,等. 铅基快堆SGTR事故下热工水力模拟及气腔扩散行为研究[J]. 原子能科学技术,2022, 56(10): 2015-2023.
[25]	辜峙钘, 余红星, 黄代顺, 等. 铅铋堆蒸汽发生器传热管破裂事故下铅铋-水相互作用程序开发及验证[J/OL]. 原子能科学技术, 2023: 1-10. （2023-04-21）[2023-05-21]. http: //kns.cnki.net/kcms/detail/11.2044.TL.20230421.0944.002.html.