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

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Abstract

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.

Details

Original languageEnglish
Title of host publicationResearch Reactor Fuel Management 10
Place of PublicationBrussels, Belgium
Publication statusPublished - Mar 2007
EventRRFM 2007 - 11th ENS Topical Meeting on Research Reactor Fuel Management - European Nuclear Society, Lyon, France
Duration: 10 Mar 200714 Mar 2007

Conference

ConferenceRRFM 2007 - 11th ENS Topical Meeting on Research Reactor Fuel Management
CountryFrance
CityLyon
Period2007-03-102007-03-14

Keywords

  • Monte Carlo burn up code MCNPX 2.6, depletion analysis

ID: 372445