A new dataset with local measurement and visualization of subcooled boiling in an internally heated annulus channel

Joseph L. Bottini, Longxiang Zhu, Zhiee Jhia Ooi, Taiyang Zhang, Caleb S. Brooks

Research output: Contribution to conferencePaper

Abstract

Thorough understanding of boiling characteristics of two-phase flow is essential for light water reactor design and safety analysis. However, detailed measurement of boiling flow is available only under limited conditions. In particular, two-phase transport data that includes local measurement of void fraction, interfacial area concentration, bubble size, and gas velocity at many axial locations is scarce in the open literature. Therefore, a new data set is obtained in an internally heated annulus channel. The inner and outer diameters of the annular channel are 19.05 mm and 38.1 mm, respectively, and the heated section is 3 m. For high axial resolution of the transport phenomena, five instrumentation ports are placed along the heated region of the internally heated annulus test section. Each instrumentation port houses two radial traversing mechanisms: one for a thermocouple to record local liquid temperature, and one for a four-point conductivity probe to measure local two-phase parameters. Viewports are also placed in the test section to capture the two-phase structure with high-speed video and to identify Onset of Nucleate Boiling (ONB) and Point of Net Vapor Generation (PNVG). The data set is widely beneficial for validation of system analysis codes and computational fluid dynamics (CFD) codes.

Original languageEnglish (US)
Pages1288-1298
Number of pages11
StatePublished - Jan 1 2019
Event18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019 - Portland, United States
Duration: Aug 18 2019Aug 23 2019

Conference

Conference18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019
CountryUnited States
CityPortland
Period8/18/198/23/19

Fingerprint

annuli
boiling
Boiling liquids
Visualization
Nucleate boiling
Light water reactors
Void fraction
Phase structure
Thermocouples
Bubbles (in fluids)
Two phase flow
light water reactors
nucleate boiling
reactor design
design analysis
Computational fluid dynamics
Systems analysis
Vapors
systems analysis
two phase flow

Keywords

  • Annulus
  • Interfacial area concentration
  • Point of net vapor generation
  • Subcooled boiling
  • Void fraction

ASJC Scopus subject areas

  • Nuclear Energy and Engineering
  • Instrumentation

Cite this

Bottini, J. L., Zhu, L., Ooi, Z. J., Zhang, T., & Brooks, C. S. (2019). A new dataset with local measurement and visualization of subcooled boiling in an internally heated annulus channel. 1288-1298. Paper presented at 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, Portland, United States.

A new dataset with local measurement and visualization of subcooled boiling in an internally heated annulus channel. / Bottini, Joseph L.; Zhu, Longxiang; Ooi, Zhiee Jhia; Zhang, Taiyang; Brooks, Caleb S.

2019. 1288-1298 Paper presented at 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, Portland, United States.

Research output: Contribution to conferencePaper

Bottini, JL, Zhu, L, Ooi, ZJ, Zhang, T & Brooks, CS 2019, 'A new dataset with local measurement and visualization of subcooled boiling in an internally heated annulus channel', Paper presented at 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, Portland, United States, 8/18/19 - 8/23/19 pp. 1288-1298.
Bottini JL, Zhu L, Ooi ZJ, Zhang T, Brooks CS. A new dataset with local measurement and visualization of subcooled boiling in an internally heated annulus channel. 2019. Paper presented at 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, Portland, United States.
Bottini, Joseph L. ; Zhu, Longxiang ; Ooi, Zhiee Jhia ; Zhang, Taiyang ; Brooks, Caleb S. / A new dataset with local measurement and visualization of subcooled boiling in an internally heated annulus channel. Paper presented at 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, Portland, United States.11 p.
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