超临界水冷堆热工水力与安全研发

赵学斌; 黄彦平; 臧金光

doi:10.13832/j.jnpe.2023.05.0223

超临界水冷堆热工水力与安全研发

doi: 10.13832/j.jnpe.2023.05.0223

中国核动力研究设计院中核核反应堆热工水力技术重点实验室，成都，610213

详细信息

作者简介:
赵学斌（1990—），男，博士，主要从事热工水力方面的研究，E-mail: xbzhao90@126.com

通讯作者:
黄彦平，E-mail: hyanping007@163.com

中图分类号: TK124；TL333
计量
- 文章访问数: 621
- HTML全文浏览量: 150
- PDF下载量: 112
- 被引次数: 0
出版历程
- 收稿日期: 2023-01-16
- 修回日期: 2023-07-25
- 刊出日期: 2023-10-13

Research and Development on Thermal Hydraulic and Safety of Supercritical Water-cooled Reactor

CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu, 610213, China

摘要

摘要: 超临界水冷堆是第四代核能系统国际论坛确定的六种先进堆型中唯一的水冷堆。由于超临界水作为冷却剂以及超临界水在物理相态的特有属性，使其在热工水力方面有着独特的表现。本文介绍了超临界水冷堆热工水力的总体要求，描述了典型热工水力过程的基本特点及目前主要研发进展，着眼于超临界水冷堆工程提出了后续研发任务，以及未来超临界水冷堆的发展建议。
- 超临界水冷堆（SCWR） /
- 热工水力 /
- 研发进展 /
- 研发任务
Abstract: Supercritical water cooled reactor (SCWR) is the only water-cooled reactor among the six advanced reactor types identified by the Generation IV International Forum. Supercritical water has a unique performance in thermal hydraulics due to its utilization as a coolant and unique properties in physical phase. This paper introduces the general requirements of SCWR thermal hydraulics, and describes the basic characteristics of typical thermal hydraulic processes and the main research and development progress at present. Focusing on the SCWR engineering application, this paper puts forward the follow-up research and development tasks and suggestions for future development of SCWR.
- SCWR /
- Thermal hydraulics /
- R&D progress /
- R&D task

HTML全文

图 1 简化堆芯多组件的并联通道示意图

Figure 1. Schematic of Simplified Parallel Channels

下载: 全尺寸图片幻灯片

图 2 SCWR 2×2棒束子通道一般划分

1—中心通道；2—角通道；3—边通道

Figure 2. General Division of Sub-channels of SCWR 2×2 Rod Bundle

下载: 全尺寸图片幻灯片

表 1 国内外超临界水临界流动实验

Table 1. Experiments on Critical Flow of Supercritical Water at Home and Abroad

实验单位	国家	实验压力 /MPa	实验段直径 /mm	实验段长径比
英国电力能源研究所^[33]	英国	22~31	1.78、2.54	1、3
中国原子能科学研究院^[34-35]	中国	22.1~29.1	1.41	3
加拿大蒙特利尔综合理工学院^[36]	加拿大	22.0~32.1	1	3

下载: 导出CSV

参考文献(45)

[1]	ZANG J G, YAN X, HUANG S F, et al. A general method for developing friction factor formulas under supercritical conditions and in different geometries[J]. Annals of Nuclear Energy, 2014, 65: 262-271. doi: 10.1016/j.anucene.2013.10.026
[2]	黄彦平, 王俊峰, 刘光旭, 等. 超临界二氧化碳热质传递与热力循环[M]. 北京: 中国原子能出版社, 2019: 93-99.
[3]	JACKSON J D. Some striking features of heat transfer with fluids at pressures and temperatures near the critical point[C]//Proceedings of the International Conference on Energy Conversion and Application. Wuhan: ICECA, 2001: 50-61.
[4]	BAE Y Y, KIM H Y. Convective heat transfer to CO₂ at a supercritical pressure flowing vertically upward in tubes and an annular channel[J]. Experimental Thermal and Fluid Science, 2009, 33(2): 329-339. doi: 10.1016/j.expthermflusci.2008.10.002
[5]	LIU S H, HUANG Y P, LIU G X, et al. Improvement of buoyancy and acceleration parameters for forced and mixed convective heat transfer to supercritical fluids flowing in vertical tubes[J]. International Journal of Heat and Mass Transfer, 2017, 106: 1144-1156. doi: 10.1016/j.ijheatmasstransfer.2016.10.093
[6]	JACKSON J D, HALL W B. Influences of buoyancy on heat transfer to fluids flowing in vertical tubes under turbulent conditions[C]. Washington: Proceedings of the Turbulent Forced Convection in Channels and Bundles, 1979: 613-640.
[7]	WANG H, BI Q C, YANG Z D, et al. Experimental and numerical investigation of heat transfer from a narrow annulus to supercritical pressure water[J]. Annals of Nuclear Energy, 2015, 80: 416-428. doi: 10.1016/j.anucene.2015.02.029
[8]	WANG H, BI Q C, WANG L C. Heat transfer characteristics of supercritical water in a 2 × 2 rod bundle - numerical simulation and experimental validation[J]. Applied Thermal Engineering, 2016, 100: 730-743. doi: 10.1016/j.applthermaleng.2016.02.083
[9]	臧金光,闫晓,黄善仿,等. 含绕丝2×2棒束内超临界水传热特性数值研究[J]. 核动力工程,2014, 35(2): 33-36.
[10]	魏佳妮,刘晓晶,柴翔,等. 7棒束超临界氟利昂流体流动传热数值分析[J]. 核科学与工程,2019, 39(6): 878-884.
[11]	WU G, WANG H, BI Q C. Experimental investigation of the pressure drop and friction factor of supercritical water in a 2×2 rod bundle[J]. Annals of Nuclear Energy, 2022, 166: 108732. doi: 10.1016/j.anucene.2021.108732
[12]	WANG H, BI Q C, WU G, et al. Experimental investigation on pressure drop of supercritical water in an annular channel[J]. The Journal of Supercritical Fluids, 2018, 131: 47-57. doi: 10.1016/j.supflu.2017.08.014
[13]	黄志刚. 竖直上升圆管内超临界水流动及传热特性研究[D]. 成都: 中国核动力研究设计院, 2011.
[14]	李玉柱, 贺五洲. 工程流体力学[M]. 北京: 清华大学出版社, 2016: 154-187.
[15]	俞冀阳. 反应堆热工水力学[M]. 第三版. 北京: 清华大学出版社, 2018: 125.
[16]	MIKHEEV M A. Fundamentals of heat transfer[M]. Moscow: Gosenergoizdat Publishing House, 1956.
[17]	POPOV V N, BELYAE V M, VALUEVA E P. Heat transfer and hydraulic drag in a turbu of helium in a circular tube at supercritical pressure[J]. High Temperature, 1978, 16: 864-871.
[18]	KIRILLOV P L, YUREV Y S, BOBKOV V P. Handbook of thermohydraulic calculations[M]. Moscow: Energoatomizdat Publishing House, 1990.
[19]	FILONENKO G K. Hydraulic resistance in pipes[J]. Teploenergetika, 1954, 1(4): 40-44.
[20]	臧金光. 棒束通道内超临界水流动及传热特性研究[D]. 成都: 中国核动力研究设计院, 2014.
[21]	XIONG T, YAN X, XIAO Z J, et al. Experimental study on flow instability in parallel channels with supercritical water[J]. Annals of Nuclear Energy, 2012, 48: 60-67. doi: 10.1016/j.anucene.2012.05.018
[22]	WANG W Y, YANG D, LIANG Z Y, et al. Experimental investigation on flow instabilities of ultra-supercritical water in parallel channels[J]. Applied Thermal Engineering, 2019, 147: 819-828. doi: 10.1016/j.applthermaleng.2018.10.107
[23]	臧金光,闫晓,黄彦平. 并联通道内超临界水流动不稳定性的数值模拟研究[J]. 核动力工程,2021, 42(2): 72-76.
[24]	CHATOORGOON V. Static instability in supercritical parallel-channel systems[C]//Proceedings of the 16th International Conference on Nuclear Engineering. Orlando: ASME, 2008: 49-56.
[25]	SU Y L, FENG J, ZHAO H, et al. Theoretical study on the flow instability of supercritical water in the parallel channels[J]. Progress in Nuclear Energy, 2013, 68: 169-176. doi: 10.1016/j.pnucene.2013.06.005
[26]	XIONG T, YAN X, HUANG S F, et al. Modeling and analysis of supercritical flow instability in parallel channels[J]. International Journal of Heat and Mass Transfer, 2013, 57(2): 549-557. doi: 10.1016/j.ijheatmasstransfer.2012.08.046
[27]	LIU G X, HUANG Y P, WANG J F, et al. Experiments on the basic behavior of supercritical CO₂ natural circulation[J]. Nuclear Engineering and Design, 2016, 300: 376-383. doi: 10.1016/j.nucengdes.2016.01.021
[28]	LIU G X, HUANG Y P, WANG J F, et al. Experimental research and theoretical analysis of flow instability in supercritical carbon dioxide natural circulation loop[J]. Applied Energy, 2017, 205: 813-821. doi: 10.1016/j.apenergy.2017.08.132
[29]	吕发, 黄彦平, 王艳林, 等. 超临界水自然循环不稳定性的实验观察[C]//中国核科学技术进展报告（第三卷）——中国核学会2013年学术年会论文集第3册（核能动力分卷（下））. 北京: 中国原子能出版社, 2013: 198-205.
[30]	YU J Y, CHE S W, LI R, et al. Analysis of Ledinegg flow instability in natural circulation at supercritical pressure[J]. Progress in Nuclear Energy, 2011, 53(6): 775-779. doi: 10.1016/j.pnucene.2011.04.001
[31]	ZHAO M F, ZHANG D X, LV Y F. A general thermal equilibrium discharge flow model[J]. Journal of Energy and Power Engineering, 2016, 10(7): 392-399.
[32]	李伟卿,张东旭,赵民富. 超临界CO₂临界流稳态试验研究及模型验证[J]. 原子能科学技术,2022, 56(8): 1593-1598.
[33]	LEE D H, SWINNERTON D. Evaluation of critical flow for supercritical steam-water: EPRINP-3086[R]. California: Electric Power Research Institute, 1983.
[34]	CHEN Y Z, ZHAO M F, YANG C S, et al. Critical flow of water under supercritical pressures[C]//Proceedings of the 14th International Heat Transfer Conference. Washington: ASME, 2010: 319-326.
[35]	CHEN Y Z, YANG C S, ZHANG S M, et al. Experimental study of critical flow of water at supercritical pressure[J]. Frontiers of Energy and Power Engineering in China, 2009, 3(2): 175-180. doi: 10.1007/s11708-009-0029-6
[36]	MUFTUOGLU A, TEYSSEDOU A. Experimental study of abrupt discharge of water at supercritical conditions[J]. Experimental Thermal and Fluid Science, 2014, 55: 12-20. doi: 10.1016/j.expthermflusci.2014.02.009
[37]	黄志刚,李永亮,曾小康,等. 竖直圆管通道内超临界水传热实验及数值模拟研究[J]. 原子能科学技术,2012, 46(7): 799-803.
[38]	HE S, KIM W S, BAE J H. Assessment of performance of turbulence models in predicting supercritical pressure heat transfer in a vertical tube[J]. International Journal of Heat and Mass Transfer, 2008, 51(19-20): 4659-4675. doi: 10.1016/j.ijheatmasstransfer.2007.12.028
[39]	张兆顺, 崔桂香, 许春晓, 等. 湍流理论与模拟[M]. 第二版. 北京: 清华大学出版社, 2017: 202.
[40]	杜代全. 超临界水冷堆子通道分析模型研究及程序开发[D]. 成都: 中国核动力研究设计院, 2012.
[41]	刘晓晶,程旭. 超临界水冷堆堆芯子通道稳态热工分析[J]. 核动力工程,2007, 28(5): 18-21,58.
[42]	许志红,杨晓,傅晟威,等. 超临界水冷堆双排燃料组件子通道分析[J]. 核科学与工程,2012, 32(1): 56-62.
[43]	RAO Y F, ONDER E N, PODILA K. Assessment of subchannel code ASSERT-PV for supercritical applications[J]. The Journal of Supercritical Fluids, 2016, 117: 164-171. doi: 10.1016/j.supflu.2016.06.016
[44]	WU P, GOU J L, SHAN J Q, et al. Preliminary safety evaluation for CSR1000 with passive safety system[J]. Annals of Nuclear Energy, 2014, 65: 390-401. doi: 10.1016/j.anucene.2013.11.031
[45]	DANG G J, LI Q, LIU Y, et al. Large-break LOCA analysis of CSR1000[J]. Annals of Nuclear Energy, 2021, 161: 108444. doi: 10.1016/j.anucene.2021.108444