基于RELAP5和FLUENT的多尺度耦合分析方法研究

doi:10.7538/yzk.2017.youxian.0824

摘要/Abstract

摘要：

基于二次开发得到的铅冷快堆一维系统程序RELAP5_LEAD和三维计算流体力学程序FLUENT，利用动态链接库技术和FLUENT用户自定义函数，开发了多尺度耦合分析程序RELAP5/FLUENT。在单相范围内，分别利用耦合程序RELAP5/FLUENT开展简单铅冷串联管道的瞬态流动和传热模拟、简单铅冷闭式回路的瞬态流动模拟，并与RELAP5_LEAD计算结果开展Code-to-Code对比分析。研究结果表明，RELAP5/FLUENT计算结果与RELAP5_LEAD模拟结果吻合良好，耦合程序的开发取得了初步成功，可用于分析铅冷快堆堆内的复杂三维热工水力现象。

关键词: 多尺度耦合, 系统程序, 计算流体力学程序, RELAP5程序, FLUENT程序, 铅冷快堆

Abstract:

Based on 1D second-development system code RELAP5_LEAD and 3D computational fluid dynamic code FLUENT, a multi-scale coupling code RELAP5/FLUENT was proposed by means of dynamic link library technique and user-defined functions of FLUENT. Under the single phase state, transient flow and heat transfer simulation in simple lead-cooled open circuit and transient flow simulation in simple lead-cooled closed circuit were selected to perform Code-to-Code verification between RELAP5/FLUENT and RELAP5_LEAD. The results show that the RELAP5/FLUENT simulation results are in good agreement with the RELAP5_LEAD calculation results and the multi-scale coupling code has achieved initial success. Through further improvement and fully validation, the multi-scale coupling code can be used to carry out complex 3D thermal-hydraulic phenomena analysis in the lead-cooled fast reactor.

Key words: multi-scale coupling, system code, computational fluid dynamic code, RELAP5 code, FLUENT code, lead-cooled fast reactor

赵鹏程;刘紫静;于涛. 基于RELAP5和FLUENT的多尺度耦合分析方法研究[J]. 原子能科学技术, 2018, 52(9): 1598-1608.

ZHAO Pengcheng;LIU Zijing;YU Tao. Research on Multi-scale Coupling Method Based on RELAP5 and FLUENT[J]. Atomic Energy Science and Technology, 2018, 52(9): 1598-1608.

参考文献

［1］CHENG X, BATTA A, BANDINI G, et al. European activities on crosscutting thermal-hydraulic phenomena for innovative nuclear systems［J］. Nuclear Engineering and Design, 2015, 290: 2-12. 
［2］ASMOLOV V G, BLINKOV V N, MELIKHOV V I, et al. Current state of system thermohydraulic codes and trends in their development abroad［J］. High Temperature, 2014, 52(1): 98-109.
［3］BOYD C. Perspectives on CFD analysis in nuclear reactor regulation［J］. Nuclear Engineering and Design, 2016, 299: 12-17.
［4］SAHA P, AKSAN N, ANDERSEN J, et al. Issues and future direction of thermal-hydraulics research and development in nuclear power reactors［J］. Nuclear Engineering and Design, 2013, 264: 3-23.
［5］BAVIERE R, TAUVERON N, PERDU F, et al. A first system/CFD coupled simulation of a complete nuclear reactor transient using CATHARE2 and TRIO_U preliminary validation on the Phénix reactor natural circulation test［J］. Nuclear Engineering and Design, 2014, 277: 124-137.
［6］彭倩,余红星,Simone,等. 反应堆热工水力中CATHARE与TRIO_U程序耦合分析研究［J］. 核动力工程,2013,34(s1):201-205.PENG Qian, YU Hongxing, Simone, et al. Analytical study on coupling of CATHARE and TRIO_U code for nuclear reactor thermal hydraulic analysis［J］. Nuclear Power Engineering, 2013, 34(s1): 201-205(in Chinese).
［7］PIALLA D, TENCHINE D, LI S, et al. Overview of the system alone and system/CFD coupled calculations of the PHENIX natural circulation test within the THINS project［J］. Nuclear Engineering and Design, 2015, 290: 78-86.
［8］BANDINI G, POLIDORI M, GERSCHENFELD A, et al. Assessment of systems codes and their coupling with CFD codes in thermal-hydraulic applications to innovative reactors［J］. Nuclear Engineering and Design, 2015, 281: 22-38.
［9］GRISHCHENKO D, JELTSOV M, KÖÖP K, et al. The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes［J］. Nuclear Engineering and Design, 2015, 290: 144-153.
［10］KÖÖP K, JELTSOV M, GRISHCHENKO D, et al. Pre-test analysis for identification of natural circulation instabilities in TALL-3D facility［J］. Nuclear Engineering and Design, 2017, 314: 110-120.
［11］PAPUKCHIEV A, LERCHL G. Extension and application of the coupled 1D-3D thermal-hydraulic code ATHLET-ANSYS CFX for the simulation of liquid metal coolant flows in advanced reactor concepts［C］∥International Conference on Nuclear Engineering and the ASME 2012 Power Conference. ［S. l.］: ［s. n.］, 2012: 2685-2694.
［12］PAPUKCHIEV A, LERCHL G, WEIS J, et al. Multiscale analysis of a transient pressurized thermal shock experiment with the coupled code ATHLET-ANSYS CFX［J］. Atw Internationale Zeitschrift fuer Kernenergie, 2012, 58(6): 402-409.
［13］PAPUKCHIEV A, JELTSOV M, KÖÖP K, et al. Comparison of different coupling CFD-STH approaches for pre-test analysis of a TALL-3D experiment［J］. Nuclear Engineering and Design, 2015, 290: 135-143.
［14］LI W, WU X, ZHANG D, et al. Preliminary study of coupling CFD code FLUENT and system code RELAP5［J］. Annals of Nuclear Energy, 2014, 73: 96-107.
［15］FENG T, ZHANG D, SONG P, et al. Numerical research on water hammer phenomenon of parallel pump-valve system by coupling FLUENT with RELAP5［J］. Annals of Nuclear Energy, 2017, 109: 318-326.
［16］冯唐涛,田文喜,宋苹,等. 基于耦合程序的流体瞬变流动水锤现象分析［J］. 原子能科学技术,2017,51(8):1364-1370.FENG Tangtao, TIAN Wenxi, SONG Ping, et al. Transient characteristics analysis of water hammer phenomena based on coupling program［J］. Atomic Energy Science and Technology, 2017, 51(8): 1364-1370(in Chinese).
［17］乔雪冬,赵勇,付陟玮. 中国实验快堆热工流体现象多维度耦合分析方法研究［J］. 核科学与工程,2012,32(3):229-233.QIAO Xuedong, ZHAO Yong, FU Zhiwei. Multi-dimension coupled simulation method of thermal-hydraulic behavior in China Experimental Fast Reactor［J］. Nuclear Science and Engineering, 2012, 32(3): 229-233(in Chinese).
［18］乔雪冬,胡文军,冯预恒,等. 中国实验快堆全厂断电事故多维度热工耦合计算［J］. 原子能科学技术,2012,46(增刊):240-245.QIAO Xuedong, HU Wenjun, FENG Yuheng, et al. Multi-dimension coupled simulation method of thermal-hydraulic behavior in China Experimental Fast Reactor under blackout accident［J］. Atomic Energy Science and Technology, 2012, 46(Suppl.): 240-245(in Chinese).
［19］PETRUZZI A, CHERUBINI M, D’AURIA F. Thirty years’ experience in RELAP5 applications at GRNSPG & NINE［J］. Nuclear Technology, 2016, 193(1): 47-87.
［20］KUMARI I, KHANNA A. Preliminary validation of RELAP5/Mod4.0 code for LBE cooled NACIE facility［J］. Nuclear Engineering and Design, 2017, 314: 217-226.
［21］BANDINI G, POLIDORI M, MELONI P, et al. RELAP5 and SIMMER-Ⅲ code assessment on CIRCE decay heat removal experiments［J］. Nuclear Engineering and Design, 2015, 281: 39-50.
［22］FAZIO C, SOBOLEV V P, AERTS A, et al. Handbook on lead-bismuth eutectic alloy and lead properties, materials compatibility, thermal hydraulics and technologies［M］. ［S. l.］: Organisation for Economic Co-Operation and Development, 2015.
［23］CHENG X, TAK N I. Investigation on turbulent heat transfer to lead-bismuth eutectic flows in circular tubes for nuclear applications［J］. Nuclear Engineering and Design, 2006, 236(4): 385-393.
［24］PFRANG W, STRUWE D. Assessment of correlations for heat transfer to the coolant for heavy liquid metal cooled core designs, FZK A7352［R］. Germany: Karlsruhe Institute of Technology, 2007.
［25］MIKITYUK K. Heat transfer to liquid metal: Review of data and correlations for tube bundles［J］. Nuclear Engineering and Design, 2009, 239(4): 680-687.
［26］JAEGER W. Empirical models for liquid metal heat transfer in the entrance region of tubes and rod bundles［J］. Heat and Mass Transfer, 2017, 53(5): 1667-1684.
［27］CHANARON B, AHNERT C, CROUZET N, et al. Advanced multi-physics simulation for reactor safety in the framework of the NURESAFE project［J］. Annals of Nuclear Energy, 2015, 84: 166-177.
［28］MYLONAKIS A G, VARVAYANNI M, CATSAROS N, et al. Multi-physics and multi-scale methods used in nuclear reactor analysis［J］. Annals of Nuclear Energy, 2014, 72: 104-119.
［29］GRUNLOH T P, MANERA A. A novel multi-scale domain overlapping CFD/STH coupling methodology for multi-dimensional flows relevant to nuclear applications［J］. Nuclear Engineering and Design, 2017, 318: 85-108.
［30］TOTI A, VIERENDEELS J, BELLONI F. Improved numerical algorithm and experimental validation of a system thermal-hydraulic/CFD coupling method for multi-scale transient simulations of pool-type reactors［J］. Annals of Nuclear Energy, 2017, 103: 36-48.
［31］CADINU F, KOZLOWSKI T, KUDINOV P. A closure-on-demand approach to the coupling of CFD and system thermal-hydraulic codes［C］∥The 7th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-7). Seoul, Korea: ［s. n.］, 2008: 269-277.

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