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Title: Effects of mesh density and flow conditioning in simulating 7-pin wire wrapped fuel pins.

Abstract

In response to the goals outlined by the U.S. Department of Energy's Global Nuclear Energy Partnership program, Argonne National Laboratory has initiated an effort to create an integrated multi-physics multi-resolution thermal hydraulic simulation tool package for the evaluation of nuclear power plant design and safety. As part of this effort, the applicability of a variety of thermal hydraulic analysis methods for the prediction of heat transfer and fluid dynamics in the wire-wrapped fuel-rod bundles found in a fast reactor core is being evaluated. The work described herein provides an initial assessment of the capabilities of the general purpose commercial computational fluid dynamics code Star-CD for the prediction of fluid dynamic characteristics in a wire wrapped fast reactor fuel assembly. A 7-pin wire wrapped fuel rod assembly based on the dimensions of fuel elements in the concept Advanced Burner Test Reactor [1] was simulated for different mesh densities and domain configurations. A model considering a single axial span of the wire wrapped fuel assembly was initially used to assess mesh resolution effects. The influence of the inflow/outflow boundary conditions on the predicted flow fields in the single-span model were then investigated through comparisons with the central span region of models whichmore » included 3 and 5 spans. The change in grid refinement had minimal impact on the inter-channel exchange within the assembly resulting in roughly a 5 percent maximum difference. The central span of the 3-span and 5-span cases exhibits much higher velocities than the single span case,, with the largest deviation (15 to 20 percent) occurring furthest away from the wire spacer grids in the higher velocity regions. However, the differences between predicted flow fields in the 3-span and 5-span models are minimal.« less

Authors:
; ; ;  [1]
  1. Mathematics and Computer Science
Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
NE
OSTI Identifier:
1014791
Report Number(s):
ANL/NE/CP-61283
TRN: US1102749
DOE Contract Number:  
DE-AC02-06CH11357
Resource Type:
Conference
Resource Relation:
Conference: 16th International Conference on Nuclear Engineering (ICONE 16); May 11, 2008 - May 15, 2008; Orlando, FL
Country of Publication:
United States
Language:
ENGLISH
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BOUNDARY CONDITIONS; BURNERS; COMPUTERIZED SIMULATION; FAST REACTORS; FLUID MECHANICS; FORECASTING; FUEL ELEMENTS; FUEL PINS; FUEL RODS; HEAT TRANSFER; NUCLEAR ENERGY; NUCLEAR ENGINEERING; NUCLEAR POWER PLANTS; RESOLUTION; SAFETY; SIMULATION; SPACERS; TEST REACTORS; THERMAL HYDRAULICS; VELOCITY

Citation Formats

Smith, J G, Babin, B R, Pointer, W D, Fischer, P F, NE), and Kansas State Univ.). Effects of mesh density and flow conditioning in simulating 7-pin wire wrapped fuel pins.. United States: N. p., 2008. Web.
Smith, J G, Babin, B R, Pointer, W D, Fischer, P F, NE), & Kansas State Univ.). Effects of mesh density and flow conditioning in simulating 7-pin wire wrapped fuel pins.. United States.
Smith, J G, Babin, B R, Pointer, W D, Fischer, P F, NE), and Kansas State Univ.). Tue . "Effects of mesh density and flow conditioning in simulating 7-pin wire wrapped fuel pins.". United States.
@article{osti_1014791,
title = {Effects of mesh density and flow conditioning in simulating 7-pin wire wrapped fuel pins.},
author = {Smith, J G and Babin, B R and Pointer, W D and Fischer, P F and NE) and Kansas State Univ.)},
abstractNote = {In response to the goals outlined by the U.S. Department of Energy's Global Nuclear Energy Partnership program, Argonne National Laboratory has initiated an effort to create an integrated multi-physics multi-resolution thermal hydraulic simulation tool package for the evaluation of nuclear power plant design and safety. As part of this effort, the applicability of a variety of thermal hydraulic analysis methods for the prediction of heat transfer and fluid dynamics in the wire-wrapped fuel-rod bundles found in a fast reactor core is being evaluated. The work described herein provides an initial assessment of the capabilities of the general purpose commercial computational fluid dynamics code Star-CD for the prediction of fluid dynamic characteristics in a wire wrapped fast reactor fuel assembly. A 7-pin wire wrapped fuel rod assembly based on the dimensions of fuel elements in the concept Advanced Burner Test Reactor [1] was simulated for different mesh densities and domain configurations. A model considering a single axial span of the wire wrapped fuel assembly was initially used to assess mesh resolution effects. The influence of the inflow/outflow boundary conditions on the predicted flow fields in the single-span model were then investigated through comparisons with the central span region of models which included 3 and 5 spans. The change in grid refinement had minimal impact on the inter-channel exchange within the assembly resulting in roughly a 5 percent maximum difference. The central span of the 3-span and 5-span cases exhibits much higher velocities than the single span case,, with the largest deviation (15 to 20 percent) occurring furthest away from the wire spacer grids in the higher velocity regions. However, the differences between predicted flow fields in the 3-span and 5-span models are minimal.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2008},
month = {1}
}

Conference:
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