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Atomic Energy

, Volume 101, Issue 2, pp 564–568 | Cite as

Validation of the MCNP5 computer code

  • E. N. Lipilina
  • V. D. Lyutov
  • É. Ya. Filippova
Article
  • 41 Downloads

Abstract

The MCNP5 computer code with the ENDF/B6 neutron data library is validated for problems which are of current importance at the Russian Federal Nuclear Center — All-Russia Research Institute of Technical Physics. Comparative calculations performed with the MCNP5 code and its preceding version MCNP4c are identical within the limits of computational error. This confirms that the MCNP5 code can be used instead of the previous versions.

Keywords

Uranium Plutonium Fuel Element Fuel Assembly Spend Fuel 
These keywords were added by machine and not by the authors. This process is experimental and the keywords may be updated as the learning algorithm improves.

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Copyright information

© Springer Science+Business Media, Inc. 2006

Authors and Affiliations

  • E. N. Lipilina
    • 1
  • V. D. Lyutov
    • 1
  • É. Ya. Filippova
    • 1
  1. 1.Russian Federal Nuclear Center — All-Russia Research Institute of Technical PhysicsRussia

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