Engineering test of HTR-PM helical tube once through steam generator
LI Xiaowei, WU Xinxin, ZHANG Zuoyi, ZHAO Jiaqing, LUO Xiaowei
Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Collaborative Innovation Center for Advanced Nuclear Energy Technology, Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084, China
摘要研发一种新型蒸汽发生器必须进行一定规模的工程验证试验。清华大学核能与新能源技术研究院在其实验基地建设了高温气冷堆示范工程(HTR-PM)螺旋管式直流蒸汽发生器的工程验证试验回路(engineering test facility for steam generator,ETF-SG)。工程验证试验回路能够模拟HTR-PM运行参数,可对HTR-PM蒸汽发生器一个换热组件进行1∶1的工程验证试验。回路设计热功率10 MW,氦回路设计压力8 MPa,最高设计温度800℃,二回路设计压力18 MPa,最高设计温度600℃。在工程验证试验回路上针对高温气冷堆蒸汽发生器完成了几十项热工水力试验,对蒸汽发生器的温度均匀性、堵管后温度分布及其温度展平调节、热工水力瞬态特性、两相流不稳定性等进行了试验研究,验证了HTR-PM螺旋管式直流蒸汽发生器的热工水力和结构设计,并为其调试、运行、低功率及启停工况参数的确定提供了重要数据和参考。
Abstract:Scale engineering tests are necessary when developing a steam generator. An engineering test facility for the steam generator (ETF-SG) was built for HTR-PM at the Institute of Nuclear and New Energy Technology (INET), Tsinghua University. ETF-SG can simulate the HTR-PM steam generator operating parameters for full scale tests of one helical tube assembly. The design thermal power is 10 MW. The design temperature and pressure of the primary helium loop are 800℃ and 8 MPa. The design temperature and pressure of the secondary loop are 600℃ and 18 MPa. Thus, ETF-SG can be used for full scale engineering tests of the test steam generator. More than twenty thermal hydraulic experiments have been finished. The tests experimentally investigated the temperature uniformity, transient thermal hydraulic response, temperature uniformity after tube plugging and adjustments, two phase flow instability and other conditions. The tests verified the thermal hydraulic and structural designs of the HTR-PM steam generator. These experiments provide important data for the commissioning and operation of the HTR-PM nuclear power plant.
李晓伟, 吴莘馨, 张作义, 赵加清, 雒晓卫. 高温气冷堆示范工程螺旋管式直流蒸汽发生器工程验证试验[J]. 清华大学学报(自然科学版), 2021, 61(4): 329-337.
LI Xiaowei, WU Xinxin, ZHANG Zuoyi, ZHAO Jiaqing, LUO Xiaowei. Engineering test of HTR-PM helical tube once through steam generator. Journal of Tsinghua University(Science and Technology), 2021, 61(4): 329-337.
[1] ZHANG Z Y, DONG Y J, LI F, et al. The Shandong Shidao Bay 200 MWe high-temperature gas-cooled reactor pebble-bed module (HTR-PM) demonstration power plant:An engineering and technological innovation[J]. Engineering, 2016, 2(1):112-118. [2] ZHANG Z Y, WU Z X, WANG D Z, et al. Current status and technical description of Chinese 2×250 MWth HTR-PM demonstration plant[J]. Nuclear Engineering and Design, 2009, 239(7):1212-1219. [3] ZHANG Z Y, SUN Y L. Economic potential of modular reactor nuclear power plants based on the Chinese HTR-PM project[J]. Nuclear Engineering and Design, 2007, 237(23):2265-2274. [4] KUGELER K, ZHANG Z Y. Modular high-temperature gas-cooled reactor power plant[M]. Beijing:Tsinghua University Press, 2019. [5] 张作义, 吴宗鑫, 王大中, 等. 我国高温气冷堆发展战略研究[J]. 中国工程科学, 2019, 21(1):12-19. ZHANG Z Y, WU Z X, WANG D Z, et al. Development strategy of high temperature gas-cooled reactor in China[J]. Strategic Study of CAE, 2019, 21(1):12-19. (in Chinese) [6] 吴宗鑫, 张作义. 世界核电发展趋势与高温气冷堆[J]. 核科学与工程, 2000, 20(3):211-219, 231. WU Z X, ZHANG Z Y. World development of nuclear power system and high temperature gas-cooled reactor[J]. Chinese Journal of Nuclear Science and Engineering, 2000, 20(3):211-219, 231. (in Chinese) [7] WU Z X, LIN D C, ZHONG D X. The design features of the HTR-10[J]. Nuclear Engineering and Design, 2002, 218(1-3):25-32. [8] 臧希年. 核电厂系统及设备[M]. 第2版. 北京:清华大学出版社, 2010. ZANG X N. Nuclear power plant systems and equipment[M]. 2nd ed. Beijing:Tsinghua University Press, 2010. (in Chinese) [9] 《蒸汽发生器》编写组. 蒸汽发生器[M]. 北京:原子能出版社, 1982. Steam Generator Team. Steam generator[M]. Beijing:Atomic Energy Press, 1982. (in Chinese) [10] STAEHLE R W. 1-Historical views on stress corrosion cracking of nickel-based alloys:The Coriou effect[M]//FERON D, STAEHLE R W. Stress corrosion cracking of nickel based alloys in water-cooled nuclear reactors, the coriou effect. Amsterdam:Elsevier Ltd., 2016. [11] GREEN S J, HETSRONI G. PWR steam generators[J]. International Journal of Multiphase Flow, 1995, 21(S1):1-97. [12] 陈长兵. 蒸汽发生器综合试验台架[J]. 中国核科技报告, 1998(S6):61. CHEN C B. Steam generator comprehensive testing facility[J]. China Nuclear Science and Technology Report, 1998(S6):61. (in Chinese) [13] 赵二雷, 李朋洲, 彭兴建, 等. 新型蒸汽发生器综合性能实验研究[J]. 原子能科学技术, 2018, 52(5):868-874. ZHAO E L, LI P Z, PENG X J, et al. Integrated experiment research of new steam generator[J]. Atomic Energy Science and Technology, 2018, 52(5):868-874. (in Chinese) [14] PROCACCIA H, DAVID J, DE PENGUERN L, et al. Thermal-hydraulic characteristics of pressurized water reactors during commercial operation Ⅱ. Steady state thermal measurements on the secondary side of a PWR steam generator (Bugey-4 Nuclear Power Plant)[J]. Nuclear Engineering and Design, 1982, 70(2):159-171. [15] WAZZAN A R, PROCACCIA H, DAVID J, et al. Thermal-hydraulic characteristics of pressurized water reactors during commercial operation:VⅡ. Thermal-hydraulics of the PWR Paluel 1 steam generator during commercial operation[J]. Nuclear Engineering and Design, 1988, 105(3):285-293. [16] DIERCKS D R, SHACK W J, MUSCARA J. Overview of steam generator tube degradation and integrity issues[J]. Nuclear Engineering and Design, 1999, 195(1):19-30. [17] YANG G Z, POINTEAU V, TEVISSEN E, et al. A review on clogging of recirculating steam generators in pressurized-water reactors[J]. Progress in Nuclear Energy, 2017, 97:182-196. [18] AUVINEN A, JOKINIEMI J K, LAHDE A, et al. Steam generator tube rupture (SGTR) scenarios[J]. Nuclear Engineering and Design, 2005, 235(2-4):457-472. [19] ANSYS Inc. ANSYS Fluent 19. 2 Theory Guide, 2019. [20] DASSAULT Inc. ABAQUS 6.13 Documentation Manual, 2019. [21] ADINA Engineering Inc. ADINA-A finite element program for automatic dynamic incremental nonlinear analysis[R]. U.S.:Watertown, 1984. [22] NRC. RELAP5/MOD3.3 code manual volume IV:Models and correlations[R]. NRC Report, NUREG/CR-5535/Rev P4, 2010. [23] NRC. TRACE V5.0 theory manual, field equations, solution methods and physical models[R]. 2008. [24] BESTION D. The physical closure laws in the CATHARE code[J]. Nuclear Engineering and Design, 1990, 124(3):229-245. [25] SINGHAL A K, SRIKANTIAH G. A review of thermal hydraulic analysis methodology for PWR steam generators and ATHOS3 code applications[J]. Progress in Nuclear Energy, 1991, 25(1):7-70. [26] Westinghouse Electric Company. GENF:A steady state performance or sizing evaluation code for model F type steam generators[R]. Rev. 4, WTD-PE-77-038, 1985. [27] BELLIARD M, GRANDOTTO M. Computation of two-phase flow in steam generator using domain decomposition and local zoom methods[J]. Nuclear Engineering and Design, 2002, 213(2-3):223-239. [28] EFFERDING L B. DYNAM:A digital computer program for study of the dynamic stability of once-through boiling flow with steam superheat[R]. GAMD-8656, 1968. [29] CHAN K C, YADIGAROGLU G. Analysis of density wave instability in counter-flow steam generators using STEAMFREQ-X[J]. Nuclear Engineering and Design, 1986, 93(1):15-24. [30] 李晓伟, 吴莘馨, 张作义. 高温气冷堆螺旋管式直流蒸汽发生器热工水力学[J]. 原子能科学技术, 2019, 53(10):1906-1917. LI X W, WU X X, ZHANG Z Y. Thermal hydraulics of HTGR helical tube once through steam generator[J]. Atomic Energy Science and Technology, 2019, 53(10):1906-1917. (in Chinese) [31] LI X W, GAO W K, SU Y, et al. Thermal analysis of HTGR helical tube once through steam generators using 1D and 2D methods[J]. Nuclear Engineering and Design, 2019, 355:110352. [32] OLSON T J, LI X W, WU X X. Tube and shell side coupled thermal analysis of an HTGR helical tube once through steam generator using porous media method[J]. Annals of Nuclear Energy, 2014, 64:67-77. [33] GAO W K, LI X W, WU X X, et al. Influences of fabrication tolerance on thermal hydraulic performance of HTGR helical tube once through steam generator[J]. Nuclear Engineering and Design, 2020, 363:110665. [34] LIANG Q, LI X W, SU Y, et al. Frequency domain analysis of two-phase flow instabilities in a helical tube once through steam generator for HTGR[J]. Applied Thermal Engineering, 2020, 168:114839. [35] MA Y, LI X W, WU X X. Thermal-hydraulic characteristics and flow instability analysis of an HTGR helical tube steam generator[J]. Annals of Nuclear Energy, 2014, 73:484-495. [36] 李晓伟. HTR-PM螺旋管式直流蒸汽发生器热工水力及温度敏感性分析[R]. 北京:清华大学核能与新能源技术研究院, 2012. LI X W. Thermal-hydraulic and temperature sensitivity analysis of the HTR-PM once through steam generator[R]. Beijing:Institute of Nuclear and New Energy Technology, 2012. (in Chinese) [37] 李晓伟. HTR-PM蒸汽发生器两相流稳定性分析报告[R]. 北京:清华大学核能与新能源技术研究院, 2018. LI X W. Two phase flow instability analysis of the HTR-PM once through steam generator[R]. Beijing:Institute of Nuclear and New Energy Technology, Tsinghua University, 2018. (in Chinese) [38] 李晓伟. HTR-PM蒸汽发生器一次侧压降核算报告[R]. 北京:清华大学核能与新能源技术研究院, 2012. LI X W. Primary pressure drop of HTR-PM helical tube once through steam generator[R]. Beijing:Institute of Nuclear and New Energy Technology, Tsinghua University, 2012. (in Chinese) [39] LI X W, ZHAO J Q, WU X X, et al. Pressure drop and heat transfer measurements of the once through steam generator helical tube bundles of high temperature gas-cooled reactors[C]//Proceedings of the 16th International Heat Transfer Conference (IHTC 16). Beijing, China, 2018. [40] ZHAO H J, LI X W, WU Y J, et al. Friction factor and Nusselt number correlations for forced convection in helical tubes[J]. International Journal of Heat and Mass Transfer, 2020, 155:119759. [41] ZHAO H J, LI X W, WU X X. New friction factor equations developed for turbulent flows in rough helical tubes[J]. International Journal of Heat and Mass Transfer, 2016, 95:525-534. [42] 张杰, 李晓伟, 吴莘馨, 等. HTR-PM蒸汽发生器入口结构对流量分配影响的数值研究[J]. 高技术通讯, 2011, 21(6):652-656. ZHANG J, LI X W, WU X X, et al. Numerical investigation of the HTR-PM steam generator entrance structure influence to the flow distribution[J]. High Technology Letters, 2011, 21(6):652-656. (in Chinese) [43] LI X W, LUO X W, WU X X, et al. Engineering verification experiments of the helical tube once through steam generator of HTR-PM[C]//Proceedings of the 2019 International Congress on Advances in Nuclear Power Plants (ICAPP2019). Juan-Les-Pins, France, 2019. [44] 李晓伟. 蒸汽发生器工程验证回路设计说明书[R]. 北京:清华大学核能与新能源技术研究院, 2012. LI X W. Specification of the engineering test facility for HTR-PM helical tube once through steam generator[R]. Beijing:Institute of Nuclear and New Energy Technology, Tsinghua University, 2012. (in Chinese) [45] 李晓伟. 高温气冷堆蒸汽发生器工程验证试验大纲[R]. 北京:清华大学核能与新能源技术研究院, 2016. LI X W. Test program for HTR-PM helical tube once through steam generator[R]. Beijing:Institute of Nuclear and New Energy Technology, Tsinghua University, 2016. (in Chinese) [46] 李晓伟, 吴莘馨, 居怀明. 高温气冷堆蒸发器工程验证实验回路蒸发器传热管流量调节特性研究[J]. 原子能科学技术, 2012, 46(S2):859-862. LI X W, WU X X, JU H M. Steam generator tube flow rate regulation of the HTGR steam generator test loop[J]. Atomic Energy Science and Technology, 2012, 46(S2):859-862. (in Chinese) [47] 李晓伟. 高温气冷堆蒸汽发生器工程验证试验报告[R]. 北京:清华大学核能与新能源技术研究院, 2019. LI X W. Experimental report for HTR-PM helical tube once through steam generator[R]. Beijing:Institute of Nuclear and New Energy Technology, Tsinghua University, 2019. (in Chinese) [48] MATHEWS A J. The early operation of the helical once-through boilers at Heysham 1 and Hartlepool[C]//Proceedings of Technology of Steam Generators for Gas-cooled Reactors-A Specialists Meeting Organized by IAEA. Winterthur, Switzerland, 1987. [49] ZHOU Y P, HAO P F, LI F, et al. Experiment study on thermal mixing performance of HTR-PM reactor outlet[J]. Nuclear Engineering and Design, 2016, 306:186-191. [50] 杨瑞昌. 热力设备水动力学讲义[M]. 北京:清华大学出版社, 2003. YANG R C. Hydrodynamics of thermal equipment[M]. Beijing:Tsinghua University Press, 2003. (in Chinese) [51] 吕俊复, 吴玉新, 李舟航, 等. 气液两相流动与沸腾传热[M]. 北京:科学出版社, 2017. LYU J F, WU Y X, LI Z H, et al. Gas liquid two phase flow and boiling heat transfer[M]. Beijing:Science Press, 2017. (in Chinese) [52] 张作义, 高祖瑛, 王大中. 两相流密度波不稳定性分析的一个显式判据[J]. 科学通报, 1990(2):146-148. ZHANG Z Y, GAO Z Y, WANG D Z. An explicit criterion for analysing twophase flow density wave instability[J]. Chinese Science Bulletin, 1990, 35(13):1129-1133. (in Chinese) [53] 张作义. 两相流不稳定性的能量原理[J]. 核科学与工程, 1989, 9(2):104-111. ZHANG Z Y. An energy principle in two phase flow instability[J]. Chinese Journal of Nuclear Science and Engineering, 1989, 9(2):104-111. (in Chinese) [54] 哈宾斯基,格尔里加. 动力装置部件中载热质的两相流动不稳定性[M]. 马昌文, 译. 北京:中国原子能出版社, 2012. HABINSKI B, GELCARY B. Two phase flow instability in power plant equipment[M]. MA C W, trans. Beijing:China Atomic Energy Press, 2012. (in Chinese) [55] 苏阳, 李晓伟, 吴莘馨. 泵驱动两相系统流动稳定性的时域理论研究[C]//中国工程热物理学会传热传质年会. 广州, 2020. SU Y, LI X W, WU X X. Time domain theoretical analysis of the two phase flow instability of pump driven two phase flow systems[C]//Proceedings of Heat and Mass Transfer Annual Meeting, Chinese Society of Engineering Thermophysics. Guangzhou, 2020. (in Chinese) [56] 苏阳, 李晓伟, 阎慧杰, 等. 物理模型及边界条件对直流蒸发管两相流不稳定性边界影响研究[J]. 原子能科学技术, 2019, 53(4):624-631. SU Y, LI X W, YAN H J, et al. Influence of physical model and boundary condition on two-phase flow instability boundary in once-through evaporation tube[J]. Atomic Energy Science and Technology, 2019, 53(4):624-631. (in Chinese)