Flow blockage analysis for the advanced neutron source reactor

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Description

The Advanced Neutron Source (ANS) reactor was designed to provide a research tool with capabilities beyond those of any existing reactors. One portion of its state-of-the-art design required high-speed fluid flow through narrow channels between the fuel plates in the core. Experience with previous reactors has shown that fuel plate damage can occur when debris becomes lodged at the entrance to these channels. Such debris disrupts the fluid flow to the plate surfaces and can prevent adequate cooling of the fuel. Preliminary ANS designs addressed this issue by providing an unheated entrance length for each fuel plate so that any ... continued below

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43 p.

Creation Information

Stovall, T.K.; Crabtree, J.A.; Felde, D.K. & Park, J.E. January 1, 1996.

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Description

The Advanced Neutron Source (ANS) reactor was designed to provide a research tool with capabilities beyond those of any existing reactors. One portion of its state-of-the-art design required high-speed fluid flow through narrow channels between the fuel plates in the core. Experience with previous reactors has shown that fuel plate damage can occur when debris becomes lodged at the entrance to these channels. Such debris disrupts the fluid flow to the plate surfaces and can prevent adequate cooling of the fuel. Preliminary ANS designs addressed this issue by providing an unheated entrance length for each fuel plate so that any flow disruption would recover, thus providing adequate heat removal from the downstream, heated portions of the fuel plates. As part of the safety analysis, the adequacy of this unheated entrance length was assessed using both analytical models and experimental measurements. The Flow Blockage Test Facility (FBTF) was designed and built to conduct experiments in an environment closely matching the ANS channel geometry. The FBTF permitted careful measurements of both heat transfer and hydraulic parameters. In addition to these experimental efforts, a thin, rectangular channel was modeled using the Fluent computational fluid dynamics computer code. The numerical results were compared with the experimental data to benchmark the hydrodynamics of the model. After this comparison, the model was extended to include those elements of the safety analysis that were difficult to measure experimentally. These elements included the high wall heat flux pattern and variable fluid properties. The results were used to determine the relationship between potential blockage sizes and the unheated entrance length required.

Physical Description

43 p.

Notes

INIS; OSTI as DE96006229

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  • Other Information: PBD: Jan 1996

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  • Other: DE96006229
  • Report No.: ORNL--6860
  • Grant Number: AC05-84OR21400
  • DOI: 10.2172/207062 | External Link
  • Office of Scientific & Technical Information Report Number: 207062
  • Archival Resource Key: ark:/67531/metadc663936

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Office of Scientific & Technical Information Technical Reports

Reports, articles and other documents harvested from the Office of Scientific and Technical Information.

Office of Scientific and Technical Information (OSTI) is the Department of Energy (DOE) office that collects, preserves, and disseminates DOE-sponsored research and development (R&D) results that are the outcomes of R&D projects or other funded activities at DOE labs and facilities nationwide and grantees at universities and other institutions.

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Creation Date

  • January 1, 1996

Added to The UNT Digital Library

  • June 29, 2015, 9:42 p.m.

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  • Jan. 22, 2016, 10:50 a.m.

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Stovall, T.K.; Crabtree, J.A.; Felde, D.K. & Park, J.E. Flow blockage analysis for the advanced neutron source reactor, report, January 1, 1996; Tennessee. (digital.library.unt.edu/ark:/67531/metadc663936/: accessed November 24, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.