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Performance of Key Features of EBR-II and the Implications for Next Generation Systems

[+] Author Affiliations
Ronald W. King, Douglas L. Porter

Argonne National Laboratory, Idaho Falls, ID

Paper No. ICONE10-22524, pp. 943-949; 7 pages
  • 10th International Conference on Nuclear Engineering
  • 10th International Conference on Nuclear Engineering, Volume 2
  • Arlington, Virginia, USA, April 14–18, 2002
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 0-7918-3596-0 | eISBN: 0-7918-3589-8
  • Copyright © 2002 by ASME


The Experimental Breeder Reactor No. 2 (EBR-II) began operation in August 1964 as an engineering test of the operation of a sodium-cooled fast reactor power plant system. Its primary mission was to demonstrate and evaluate the performance of a sodium-cooled reactor system using recycled fuel from an integral fuel cycle facility, and as an electrical power generator tied to the utility grid. It accomplished this mission in the early years of operation. Following the early successes, the mission evolved to include more extensive testing of fuels, materials, components, and safety features of a sodium-cooled fast reactor. The use of sodium as the coolant, use of metal fuel, and a piped-pool primary system configuration, were key contributors to its notable long-term performance. Extensive evaluation and examination of these features have provided a solid basis for and understanding of the technology. Recent interest in future designs for nuclear generating stations has generated renewed interest in liquid-metal-cooled fast reactors in the United States and elsewhere. The successful operation of EBR-II for thirty years and the demonstration of characteristics significant to the development of next-generation reactors, has prompted a re-examination of key features of EBR-II and a review of its performance record. This paper discusses selected key features and their contribution to the performance of EBR-II, evaluates the overall performance of the reactor, and discusses the implications for the development of next generation reactor concepts.

Copyright © 2002 by ASME



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