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

Hui Tianyu; Tong Lili; Cao Xuewu

doi:10.11884/HPLPB202436.230408

Volume 36 Issue 7

May 2024

Turn off MathJax

Article Contents

Article Navigation > High Power Laser and Particle Beams > 2024 > 36(7): 076002

Hui Tianyu, Tong Lili, Cao Xuewu. Simulation of coolant boiling phenomenon in sodium cooled fast reactor based on porous medium approach[J]. High Power Laser and Particle Beams, 2024, 36: 076002. doi: 10.11884/HPLPB202436.230408

Citation:

Hui Tianyu, Tong Lili, Cao Xuewu. Simulation of coolant boiling phenomenon in sodium cooled fast reactor based on porous medium approach[J]. High Power Laser and Particle Beams, 2024, 36: 076002. doi: 10.11884/HPLPB202436.230408

Citation:

PDF( 1317 KB)

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

doi: 10.11884/HPLPB202436.230408

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

Received Date: 2023-11-27
Accepted Date: 2024-03-01
Rev Recd Date: 2024-03-01

Available Online: 2024-03-20

Publish Date: 2024-05-31

Abstract

Abstract

Accurate prediction of the occurrence time and location of coolant boiling is of great significance for 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 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 s, which is 0.2 s different from the result of experiment. The overall trend of temperature and flow rate changes are in good agreement with the experimental data.
- sodium cooled fast reactor,
- sodium boiling,
- porous medium analysis approach,
- two-fluid six-equation model

FullText(HTML)

References(28)

References

[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