Computational fluid dynamics prediction of grid spacer thermal-hydraulic performance with comparison to experimental results

Robert Lee Campbell, John Michael Cimbala, Lawrence E. Hochreiter

Research output: Contribution to journalArticle

3 Citations (Scopus)

Abstract

The thermal-hydraulic performance of a nuclear reactor fuel assembly grid spacer is predicted using computational fluid dynamics. The modeled flow domain exploits the periodicity of the spacer and is separated into a bare bundle and grid region to maintain a manageable model size. An iterative process is used to couple the segregated flow domains to arrive at a converged solution. The grid spacer is a 7 × 7 mixing vane grid representative of an actual pressurized water reactor grid. Pressure drop and rod wall temperature predictions for steady-state operation are computed. The results show excellent agreement with experimental data. The agreement in these results demonstrates the usefulness of the method presented as a design tool for nuclear fuel manufacturers and as a prediction tool for off-design operating conditions such as simulated accident scenarios.

Original languageEnglish (US)
Pages (from-to)49-61
Number of pages13
JournalNuclear Technology
Volume149
Issue number1
DOIs
StatePublished - Jan 1 2005

Fingerprint

computational fluid dynamics
hydraulics
spacers
Computational fluid dynamics
grids
Hydraulics
nuclear fuels
Pressurized water reactors
Nuclear fuels
Nuclear reactors
predictions
Pressure drop
Accidents
pressurized water reactors
vanes
nuclear reactors
wall temperature
accidents
pressure drop
bundles

All Science Journal Classification (ASJC) codes

  • Nuclear and High Energy Physics
  • Nuclear Energy and Engineering
  • Condensed Matter Physics

Cite this

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abstract = "The thermal-hydraulic performance of a nuclear reactor fuel assembly grid spacer is predicted using computational fluid dynamics. The modeled flow domain exploits the periodicity of the spacer and is separated into a bare bundle and grid region to maintain a manageable model size. An iterative process is used to couple the segregated flow domains to arrive at a converged solution. The grid spacer is a 7 × 7 mixing vane grid representative of an actual pressurized water reactor grid. Pressure drop and rod wall temperature predictions for steady-state operation are computed. The results show excellent agreement with experimental data. The agreement in these results demonstrates the usefulness of the method presented as a design tool for nuclear fuel manufacturers and as a prediction tool for off-design operating conditions such as simulated accident scenarios.",
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Computational fluid dynamics prediction of grid spacer thermal-hydraulic performance with comparison to experimental results. / Campbell, Robert Lee; Cimbala, John Michael; Hochreiter, Lawrence E.

In: Nuclear Technology, Vol. 149, No. 1, 01.01.2005, p. 49-61.

Research output: Contribution to journalArticle

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