Differences in experimental friction factors across two model helical coil steam generators

M. Delgado, S. Lee, Y. A. Hassan

Research output: Contribution to conferencePaperpeer-review

2 Scopus citations

Abstract

Tube and shell heat exchanger geometries have been extensively studied for their increased heat transfer properties within a smaller region than typical heat exchangers. Current designs of nuclear power plants have incorporated a particular tube and shell model, the helical coil steam generator for its economical and efficient design. Characteristic properties include multiple layers of tube bundles that coil against one another at different pitches. The complex helical coil steam generator design makes it difficult for traditional computational methods to model fluidic properties of the external flow. Two experimental test facilities have been built from a helical coil steam generator configuration to study the external flow friction factor across a series of tubes bundles. The closed loop flow facilities were designed to control the flow velocity at the inlet of the experimental test section. Experiments to measure pressure drop were conducted on both test facilities with different number of tube bundles between them. The first test section modeled a single channel flow between two coiled interfacing rod bundles of twelve rods. The second test section modeled multiple channel flow across a test section of a five coiled tube bundle assembly. This assembly is composed of three bundles of eighteen rods that coil against two bundles of nine rods in between them. Pressure drop data was acquired for different Re ranges on both facilities with a maximum of approximately 120,000. Data results between the model helical coil steam generators were normalized and compared. Friction factor calculated from the fluidic properties measured were then compared to several existing models of tube and shell heat exchangers.

Original languageEnglish (US)
StatePublished - 2017
Event17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017 - Xi'an, Shaanxi, China
Duration: Sep 3 2017Sep 8 2017

Other

Other17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017
Country/TerritoryChina
CityXi'an, Shaanxi
Period9/3/179/8/17

All Science Journal Classification (ASJC) codes

  • Nuclear Energy and Engineering
  • Instrumentation

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