期刊文献+

稠密带缠绕丝棒束阻力特性研究 被引量:2

Investigation on Flow Resistance Coefficient of Wires Wrapped Tight Lattice Rods Bundle
下载PDF
导出
摘要 为获得稠密布置燃料组件的阻力系数,应用稠密带缠绕丝棒束进行实验研究,拟合阻力系数关系式,并将关系式与经典Rehme关系式进行比较分析。结果表明Rehme关系式不适用于本实验棒束。同时应用计算流体力学(CFD)方法、剪切应力输运模型(SST)湍流模型对实验进行模拟,获得棒束内部的流动形式、压力场和沿程阻力系数,并与实验结果进行对比。结果表明CFD方法可作为预测稠密带缠绕丝棒束单相流动阻力系数的参考。 为获得稠密布置燃料组件的阻力系数,应用稠密带缠绕丝棒束进行实验研究,拟合阻力系数关系式,并将关系式与经典Rehme关系式进行比较分析。结果表明Rehme关系式不适用于本实验棒束。同时应用计算流体力学(CFD)方法、剪切应力输运模型(SST)湍流模型对实验进行模拟,获得棒束内部的流动形式、压力场和沿程阻力系数,并与实验结果进行对比。结果表明CFD方法可作为预测稠密带缠绕丝棒束单相流动阻力系数的参考。
出处 《核动力工程》 EI CAS CSCD 北大核心 2012年第S1期45-49,共5页 Nuclear Power Engineering
关键词 稠密棒束 绕丝 阻力特性 实验研究 计算流体力学 Tight lattice Wires wrapped Resistance coefficient Experimental study CFD
  • 相关文献

参考文献10

  • 1Imteyaz Ahmad,Kwang-Yong Kim.Flow and convective heat transfer analysis using RANS for a wire-wrapped fuel assembly[J]. Journal of Mechanical Science and Technology . 2006 (9)
  • 2GAJAPATHY R,VELUSAMY K,SELVARAJ P,et al.CFD investigation of helical wire-wrapped7-pin fuel bundle and the challenges in modeling full scale217pin bundle. Nuclear Engineer The . 2007
  • 3Oldekop W,Berger H D,Zeggel W.General features of advanced pressurized water reactors with improved fuel utilization. Nuclear Technology . 1982
  • 4Zeggel W.CHF in the Parameter Range of AdvancedPWR Cores. Nuclear Engineer The . 1987
  • 5Zeggel W.Status of Tight-Lattice Thermal-Hydraulics. IAEA-Metting on Technical and Economic Aspectsof High Converters . 1990
  • 6Gambier G,Lombardi P.Development Trends for FutureFrench Pressurized Water Reactors. Nuclear Techno-logy . 1988
  • 7Reheme K.Pressure Drop Correlations for Fuel ElementSpacers. Nuclear Technology . 1973
  • 8Reheme K.Pressure Drop Performance of Rod Boundlessin Hexagonal Arrangements. International Journal of Heat and Mass Transfer . 1972
  • 9Wasim Raza,Kwang-Yong Kim.Comparative Analysisof Flow and Convective Heat Transfer between 7-Pin and19-Pin Wire-Wrapped Fuel Assemblies. NuclearScience and Technology . 2008
  • 10Chandra L,Roelofs F.CFD Analyses on the Influence ofWire Wrap Spacers on Heat Transfer at SupercriticalConditions. Proc.4th Int.Symposium on SCWR . 2009

同被引文献12

  • 1Takizuka T, Tsujimoto K, Sasa T, et al.. Design study of lead-bismuth cooled ADS dedicated to nuclear waste tranmutataion [J]. Prog. Nucl. Energy 40(3-4): 505-.512.
  • 2OECD/NEA. Handbook on lead-bismuth eutectic alloy and lead properties, materials compatibility, Thermal-hydraulics and Technologies[R]. ISBN978- 92-64-99002-9, OECD (2007).
  • 3Bubelis E. Review and roposal for best fit of wire- wrapped fuel bundle Friction factor and pressure Drop Predictions Using Various Existing correlation[J]. Nucl. Eng. Des. 238: 3299-3320.
  • 4Wu Y, Bai Y, Song Y, et al. Overview of lead-bismuth reactor design and R&D status in china [C]. The 13th International Conference on Fast Reactors and Related Cycles: Safe Technologies and Sustainable Scenarios, Paris, France. 4-7 March 2013. Novendstern E H, 1972.
  • 5Turbulent flow pressure drop model for fuel rod assemblies utilizing a helical wire-wrap spacer system[J]. Nucl. Eng. Des. 22: 19-27.
  • 6Rehme K, 1972. Pressure drop correlations for fuel element spacers [J]. Nucl. Technol. 17: 15-23.
  • 7于意奇,顾汉洋,杨燕华,程旭,王小军.紧密栅元棒束通道内的传热流动数值模拟[J].核动力工程,2011,32(2):53-58. 被引量:3
  • 8詹文龙,徐瑚珊.未来先进核裂变能——ADS嬗变系统[J].中国科学院院刊,2012,27(3):375-381. 被引量:145
  • 9靖剑平,张春明,陈妍,孙微,庄少欣.浅谈核电领域中的热工水力分析程序[J].核安全,2012,11(3):70-74. 被引量:19
  • 10曹红军,闫修平.首台AP1000-回路水压试验方案及风险分析[J].核安全,2013,12(2):39-44. 被引量:6

引证文献2

二级引证文献3

相关作者

内容加载中请稍等...

相关机构

内容加载中请稍等...

相关主题

内容加载中请稍等...

浏览历史

内容加载中请稍等...
;
使用帮助 返回顶部