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.
|Title of host publication||Research Reactor Fuel Management 10|
|Place of Publication||Brussels, Belgium|
|State||Published - Mar 2007|
|Event||2007 - RRFM - Research Reactor Fuel Management: 11th ENS Topical Meeting - European Nuclear Society, Lyon|
Duration: 10 Mar 2007 → 14 Mar 2007
|Conference||2007 - RRFM - Research Reactor Fuel Management|
|Period||2007-03-10 → 2007-03-14|