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Home / Publications / Journals / Nuclear Technology / Volume 164 / Number 1 / Pages 103-118

Temperature and Flow Distributions in Sodium-Heated Large Straight Tube Steam Generator by Numerical Methods

Naoyuki Kisohara, Takeshi Moribe, Takaaki Sakai

Nuclear Technology / Volume 164 / Number 1 / October 2008 / Pages 103-118

Technical Paper / Icapp '06

A sodium-heated steam generator (SG) being studied in Japan for a future commercialized fast reactor is a double-wall straight tube type. The SG is large to reduce its manufacturing cost by economies of scale. This paper addresses the multidimensional distributions of the temperature and the flow in the SG. Large heat exchanger components are prone to have nonuniform flow and temperature distributions. These maldistributions cause tubes to have thermal expansion mismatch, which might lead to structural issues such as tube buckling or tube-to-tube-sheet junction failure in straight tube SGs. The temperature profiles in the SG are examined by numerical methods, and flow distribution control devices are optimized to prevent these issues. The calculation model of the SG consists of two parts: a sodium inlet distribution plenum (the inlet plenum) and a heat transfer tube bundle region (the bundle). The flow and temperature distributions in the inlet plenum and the bundle are evaluated by the three-dimensional code FLUENT and the two-dimensional code MSG, respectively. The thermal loads on the tubes are evaluated by the structural code FINAS based on the temperature distributions. These codes have revealed that the sodium flow is distributed uniformly by the flow distributors and that the thermal loads remain within the allowable range for the structural integrity of the tubes and the junctions. An inlet plenum water test and an SG experiment to examine thermal-hydraulic performance are planned. These tests will reveal the flow and temperature distributions in the SG and verify the computer calculation results.

 
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