基于多孔介质方法的钠冷快堆冷却剂沸腾现象模拟

惠天宇; 佟立丽; 曹学武

doi:10.11884/HPLPB202436.230408

基于多孔介质方法的钠冷快堆冷却剂沸腾现象模拟

doi: 10.11884/HPLPB202436.230408

上海交通大学机械与动力工程学院，上海 200240

基金项目: 国家自然科学基金项目(U1967202)

详细信息

作者简介:
惠天宇，hty1998@sjtu.edu.cn

通讯作者:
佟立丽，lltong@sjtu.edu.cn。

中图分类号: TL33
计量
- 文章访问数: 25
- HTML全文浏览量: 20
- PDF下载量: 3
- 被引次数: 0
出版历程
- 收稿日期: 2023-11-27
- 修回日期: 2024-03-01
- 录用日期: 2024-02-06
- 网络出版日期: 2024-03-20

Simulation of coolant boiling phenomenon in sodium cooled fast reactor based on porous medium approach

School of Mechanical Engineering, Shanghai Jiao Tong University, Shanghai 200240, China

摘要

摘要: 基于两流体六方程模型针对钠的气液两相分别构建守恒方程，采用蒸发冷凝模型表征两相质量交换，分别使用显式和隐式处理方法对蒸发冷凝模型进行计算，同时考虑了Sobolev阻力模型、两相对流换热模型以及相间动量交换等本构关系，开发了适用于模拟钠冷快堆冷却剂沸腾的多孔介质分析方法，利用KNS-37失流实验L22工况数据进行了对比验证，并利用L29工况流量数据验证模型的适用性。结果表明，所建立的钠沸腾多孔介质分析方法可以对钠冷快堆沸腾现象较好地模拟，预测沸腾发生时间在6.3 s左右，与实验相差0.2 s，温度和流量的总体变化趋势与实验数据吻合较好。
- 钠冷快堆 /
- 钠沸腾 /
- 多孔介质方法 /
- 两流体六方程
Abstract: As the first stage of severe accidents in sodium cooled fast reactors, accurate prediction of the occurrence time and location of coolant boiling is of great significance for the safety assessment of Sodium Cooled Fast Reactors (SFR). Based on a two fluid six equation model, conservation equations are constructed for the gas-liquid two-phase flow of sodium. The evaporation-condensation model is used to characterize the interphase mass exchange, and explicit and implicit methods are used to calculate evaporation-condensation model. Constitutive relationships such as Sobolev resistance model, two phase flow heat transfer model, and phase momentum exchange are considered. A porous medium analysis approach which is suitable for simulating SFR coolant boiling was developed, and comparative verification was conducted using KNS-37 L22 loss of flow experiment data. L29 flow data is used to verify the applicability of the model. The results indicate that the established sodium boiling porous medium analysis approach can effectively simulate the boiling phenomenon. It predicts that the boiling time will be around 6.3 seconds, which is 0.2 seconds different from the experiment. The overall trend of temperature and flow rate changes are in good agreement with experimental data.
- Sodium Cooled Fast Reactors /
- sodium boiling /
- porous medium analysis approach /
- two-fluid six-equation model

HTML全文

图 1 基于多孔介质方法的网格划分

Figure 1. Mesh generation based on porous medium approach

下载: 全尺寸图片幻灯片

图 2 KNS-37钠沸腾回路示意图

Figure 2. KNS-37 sodium boiling loop

下载: 全尺寸图片幻灯片

图 3 通道划分示意图

Figure 3. Schematic diagram of channel division

下载: 全尺寸图片幻灯片

图 4 沸腾前不同时刻中心通道冷却剂温度轴向分布

Figure 4. Axial profile of central channel coolant temperature at different times before boiling

下载: 全尺寸图片幻灯片

图 5 工况L22质量流量变化

Figure 5. Mass flow trend of L22

下载: 全尺寸图片幻灯片

图 6 工况L22中心通道温度及压力变化

Figure 6. Temperature and pressure in central channel of L22

下载: 全尺寸图片幻灯片

图 7 工况L29与L22进口流量变化

Figure 7. Inlet mass flow trend of L29 and L22

下载: 全尺寸图片幻灯片

表 1 KNS-37 L22实验参数

Table 1. Experimental parameters of KNS-37 test L22

experimental parameters	inlet pressure/ bar	outlet pressure/ bar	inlet temperature/ ℃	outlet temperature/ ℃	inlet mass flow rate/kg·s⁻¹	average line power density/W·cm⁻¹
value	2.241	1.045	380	539	3.41	215.4

下载: 导出CSV

表 2 沸腾时间与干涸时间预测结果

Table 2. Prediction results of boiling time and drying time

	Boiling onset/ s	Dry-out onset/ s
experiment	6.11	9.25
model 1	6.6	7.0
model 2	6.3	10.7
TRACE	5.73	9.16
NATOF-2D	6.30	6.40
BACCHUS	6.28	9.2
SABENA	6.20	9.27

下载: 导出CSV

[1]	Tentner A M, Parma E, Wei T, et al. Severe accident approach-final report. Evaluation of design measures for severe accident prevention and consequence mitigation[R]. Argonne: Argonne National Laboratory, 2010.
[2]	Bachrata A, Bertrand F, Marie N, et al. A Comparative study on severe accident phenomena related to melt progression in Sodium Fast Reactors and Pressurized Water Reactors[J]. Journal of Nuclear Engineering and Radiation Science, 2021, 7: 030801. doi: 10.1115/1.4047921
[3]	周科源, 喻宏, 胡赟, 等. CEFR钠空泡反应性效应试验测量与计算分析[J]. 原子能科学技术, 2013, 47(s1):70-74 doi: 10.7538/yzk.2013.47.zk.0070 Zhou Keyuan, Yu Hong, Hu Yun, et al. Measurement and analysis of CEFR sodium void reactivity effect[J]. Atomic Energy Science and Technology, 2013, 47(s1): 70-74 doi: 10.7538/yzk.2013.47.zk.0070
[4]	Waltar A E, Todd D R, Tsvetkov P V. Fast spectrum reactors[M]. New York: Springer, 2012.
[5]	Končar B, Matkovičc M, Prošek A. NEPTUNE_CFD analysis of flow field in rectangular boiling channel[J]. The Journal of Computational Multiphase Flows, 2012, 4(4): 399-409. doi: 10.1260/1757-482X.4.4.399
[6]	Mimouni S, Guingo M, Lavieville J. Assessment of RANS at low Prandtl number and simulation of sodium boiling flows with a CMFD code[J]. Nuclear Engineering and Design, 2017, 312: 294-302. doi: 10.1016/j.nucengdes.2016.07.006
[7]	Rose S D, Dearing J F. Post-test analysis of dryout test 7B' of the W-1 sodium loop safety facility experiment with the SABRE-2P code[R]. Milwaukee: Oak Ridge National Laboratory, 1981.
[8]	Ninokata H, Okano T A. Sabena: Subassembly boiling evolution numerical analysis[J]. Nuclear Engineering and Design, 1990, 120(2/3): 349-367.
[9]	No H C, Kazimi M S. An investigation of the physical foundations of two-fluid representation of sodium boiling in the liquid-metal fast breeder reactor[J]. Nuclear Science and Engineering, 1987, 97(4): 327-343. doi: 10.13182/NSE87-A23516
[10]	Kruessmann R, Ponomarev A, Pfrang W, et al. Assessment of SFR reactor safety issues: Part II: Analysis results of ULOF transients imposed on a variety of different innovative core designs with SAS-SFR[J]. Nuclear Engineering and Design, 2015, 285: 263-283. doi: 10.1016/j.nucengdes.2014.11.037
[11]	Perez-Martin S, Pfrang W, Haselbauer M. Analysis of the CABRI-1 single fuel pin LOF experiment BI1 with SAS-SFR code including two-phase sodium behavior[C]//Proceedings of the ICAPP 2014. 2014.
[12]	Chenu A, Mikityuk K, Chawla R. TRACE simulation of sodium boiling in pin bundle experiments under loss-of-flow conditions[J]. Nuclear Engineering and Design, 2009, 239(11): 2417-2429. doi: 10.1016/j.nucengdes.2009.07.015
[13]	吴宗芸, 刘天才, 吴明宇. 基于双区域模型的钠冷快堆组件子通道分析程序的开发与验证[J]. 原子能科学技术, 2022, 56(4):672-683 doi: 10.7538/yzk.2021.youxian.0297 Wu Zongyun, Liu Tiancai, Wu Mingyu. Development and validation of subchannel analysis program based on two-region model for sodium cooled fast reactor assembly[J]. Atomic Energy Science and Technology, 2022, 56(4): 672-683 doi: 10.7538/yzk.2021.youxian.0297
[14]	方闻韬, 佟立丽, 曹学武. 绕丝交混模型对钠冷快堆组件子通道分析的影响[J]. 强激光与粒子束, 2023, 35:096001 doi: 10.11884/HPLPB202335.230051 Fang Wentao, Tong Lili, Cao Xuewu. Influence of wire wrap mixing model on sub-channel analysis of sodium-cooled fast reactor assembly[J]. High Power Laser and Particle Beams, 2023, 35: 096001 doi: 10.11884/HPLPB202335.230051
[15]	Grand D, Basque G. Two-dimensional calculation of sodium boiling in sub-assemblies[C]//International Meeting on Fast Reactor Safety Technology. 1979.
[16]	Fukano Y. SAS4A analysis on hypothetical total instantaneous flow blockage in SFRs based on in-pile experiments[J]. Annals of Nuclear Energy, 2015, 77: 376-392. doi: 10.1016/j.anucene.2014.11.034
[17]	Domanus H M, Shah V L, Sha W T. Applications of the COMMIX code using the porous medium formulation[J]. Nuclear Engineering and Design, 1980, 62(1/3): 81-100.
[18]	Schor A L, Kazimi M S, Todreas N E. Advances in two-phase flow modeling for LMFBR applications[J]. Nuclear Engineering and Design, 1984, 82(2/3): 127-155.
[19]	徐迟, 李文龙, 谢淳, 等. 金属钠蒸发行为研究[J]. 原子能科学技术, 2022, 56(3):467-474 doi: 10.7538/yzk.2021.youxian.0792 Xu Chi, Li Wenlong, Xie Chun, et al. Evaporation behavior of metal sodium[J]. Atomic Energy Science and Technology, 2022, 56(3): 467-474 doi: 10.7538/yzk.2021.youxian.0792
[20]	Silver R S, Simpson H C. The condensation of superheated steam[C]//Proceeding of National England Laboratory Conference. 1961.
[21]	Schrage R W. A theoretical study of interphase mass transfer[M]. New York: Columbia University Press, 1953.
[22]	Sobolev V. Fuel rod and assembly proposal for XT-ADS pre-design[C]//Coordination meeting of WP1&WP2 of DM1 IP EUROTRANS. 2006: 8-9.
[23]	Chen J C. A correlation for boiling heat transfer in convection flow[R]. United States, 1962.
[24]	Borishanskii V M, Gotovskii M A, Firsova É V. Heat transfer to liquid metals in longitudinally wetted bundles of rods[J]. Soviet Atomic Energy, 1969, 27(6): 1347-1350. doi: 10.1007/BF01118660
[25]	Dittus F W, Boelter L M K. Heat transfer in automobile radiators of the tubular type[J]. International Communications in Heat and Mass Transfer, 1985, 12(1): 3-22. doi: 10.1016/0735-1933(85)90003-X
[26]	Bottoni M, Dorr B, Homann C, et al. Experimental and numerical investigations of sodium boiling experiments in pin bundle geometry[J]. Nuclear Technology, 1990, 89(1): 56-82. doi: 10.13182/NT90-A34359
[27]	Huber F, Kaiser A, Mattes K, et al. Steady state and transient sodium boiling experiments in a 37-pin bundle[J]. Nuclear Engineering and Design, 1987, 100(3): 377-386. doi: 10.1016/0029-5493(87)90087-2
[28]	Perez-Martin S, Anderhuber M, Laborde L, et al. Evaluation of sodium boiling models using KNS-37 loss of flow experiments[J]. Journal of Nuclear Engineering and Radiation Science, 2022, 8: 011310. doi: 10.1115/1.4050769