MCNPX 2.6.C vs. MCNPX & ORIGEN-S – State of the Art for Reactor Core Management

Silva Kalcheva, Edgar Koonen

    Research outputpeer-review


    This paper discusses the application of the Monte Carlo burn up code MCNPX 2.6.C for the criticality and depletion reactor core analysis of the Material Testing Research Reactor BR2 in SCK•CEN in Mol, Belgium. A comparison with the developed at the BR2 reactor department combined MCNP&ORIGEN-S fuel depletion method is presented. The accuracy of the both methods, the consumption of the calculation time, the depletion capabilities, the advantages and disadvantages of use of the both methods are discussed. Validation of MCNPX 2.6.C is performed on the reactivity measurements at the Reactor BR2.
    Original languageEnglish
    Title of host publicationResearch Reactor Fuel Management 10
    Place of PublicationBrussels, Belgium
    StatePublished - Mar 2007
    Event2007 - RRFM - Research Reactor Fuel Management: 11th ENS Topical Meeting - European Nuclear Society, Lyon
    Duration: 10 Mar 200714 Mar 2007


    Conference2007 - RRFM - Research Reactor Fuel Management

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