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Thermal-hydraulic performance of heavy liquid metal in straight-tube and U-tube heat exchangers

机译:重金属在直管式和U型管式换热器中的热工液压性能

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摘要

Motivated by an increased interest in heavy liquid metal (lead or lead alloy) cooled fast reactors (LFR) and accelerator-driven system (ADS), the present paper presents a study on resistance characteristics and heat transfer performance of liquid lead bismuth eutectic (LBE) flow through a straight-tube heat exchanger and a U-tube heat exchanger. The investigation is performed on the TALL test facility at KTH. The heat exchangers have counter-current flow arrangement, and are made from a pair of 1-m-long concentric ducts, with the LBE flowing in the inner tube of 10 mm I.D. and the secondary coolant flowing in the annulus. The inlet temperature of LBE into the heat exchangers is from 200 ℃ to 450 ℃ with temperature drops from 0℃ to 100℃ within the LBE flow range of Re = 10~4-10~5. Analysis of the experimental results obtained provides a basic understanding and quantification of the regimes of lead-bismuth flow and heat transfer through a straight tube and a U-shaped tube. The unique data base also serves as benchmark and improvement for system thermal-hydraulic codes (e.g. RELAP, TRAC/AAA) whose development and testing were dominantly driven by applications in water-cooled systems. Lessons and insights learnt from the study and recommendations for the heat exchanger selection are discussed.
机译:出于对重金属液态(铅或铅合金)冷却快堆(LFR)和加速器驱动系统(ADS)的兴趣增加,本文提出了对液态铅铋共晶(LBE)的电阻特性和传热性能的研究。 )流过直管式换热器和U型管式换热器。该调查是在KTH的TALL测试设施上进行的。热交换器具有逆流布置,由一对1米长的同心导管制成,LBE在10 mm I.D的内管中流动。二次冷却剂在环空中流动。在Re = 10〜4-10〜5的LBE流量范围内,LBE进入热交换器的温度为200℃至450℃,温度从0℃下降至100℃。对获得的实验结果的分析提供了对铅-铋流动和通过直管和U形管的传热机制的基本理解和量化。独特的数据库还可以用作系统热工代码(例如RELAP,TRAC / AAA)的基准和改进,其开发和测试主要是由水冷系统中的应用驱动的。讨论了从研究中吸取的教训和见解,以及对换热器选择的建议。

著录项

  • 来源
    《Nuclear Engineering and Design》 |2009年第7期|1323-1330|共8页
  • 作者单位

    Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21,106 91 Stockholm, Sweden;

    Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21,106 91 Stockholm, Sweden;

    Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21,106 91 Stockholm, Sweden;

    Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21,106 91 Stockholm, Sweden;

  • 收录信息 美国《科学引文索引》(SCI);美国《工程索引》(EI);美国《生物学医学文摘》(MEDLINE);
  • 原文格式 PDF
  • 正文语种 eng
  • 中图分类
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