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Thermal-hydraulic performance of lead-bismuth eutectic in a straight-tube and a u-tube heat exchangers
KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
2007 (English)In: Proceedings of the 12th International Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12), 2007Conference paper, Published paper (Refereed)
Abstract [en]

Motivated by an increased interest in 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 molten 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-meter-long concentric ducts, with the LBE flowing in the inner tube of 10mm ID and the secondary coolant flowing in the annulus. The inlet temperature of LBE into the heat exchangers is from 200°C to 450°C with temperature drops from 0°C to 100°C within the LBE flow range of Re=10 4-105. 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.

Place, publisher, year, edition, pages
2007.
Keyword [en]
Fast reactor, Flow resistance, Heat transfer, Heavy liquid metal
National Category
Energy Engineering
Identifiers
URN: urn:nbn:se:kth:diva-53781Scopus ID: 2-s2.0-44349134568ISBN: 0-89448-058-8 (print)ISBN: 978-089448058-4 OAI: oai:DiVA.org:kth-53781DiVA: diva2:470959
Conference
12th International Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12) Pittsburgh, Pennsylvania, September 30-October 4, 2007
Note
QC 20120113Available from: 2011-12-30 Created: 2011-12-30 Last updated: 2012-01-13Bibliographically approved

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