Application of MCNPX 2.7.D for Reactor Core Management at the Research Reactor BR2

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Abstract

The paper discusses application of the Monte Carlo burn up code MCNPX 2.7.D for whole core criticality and depletion analysis of the Material Testing Research Reactor BR2 at SCK•CEN in Mol, Belgium. Two different approaches in the use of MCNPX 2.7.D are presented. The first methodology couples the evolution of fuel depletion, evaluated by MCNPX 2.7.D in an infinite lattice with a steady-state 3-D power distribution in the full-scale core model. The second method represents fully automatic whole core depletion and criticality calculations in the full-scale 3-D heterogeneous geometry model of the BR2 reactor. The accuracy of the method and computational time as function of the number of used unique burn up materials in the model are being studied. The depletion capabilities of MCNPX 2.7.D are compared with the MCNPX & ORIGEN-S combined method developed at the BR2 reactor department. Testing on criticality measurements at the BR2 reactor is presented.

Details

Original languageEnglish
Title of host publicationInternational Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2011)
Place of PublicationLaGrange Park, IL, United States
Publication statusPublished - 18 May 2011
EventMC 2011 - Rio de Janeiro, Brazil
Duration: 8 May 201112 May 2011

Conference

ConferenceMC 2011
Country/TerritoryBrazil
CityRio de Janeiro
Period2011-05-082011-05-12

Keywords

  • MCNPX 2.7.D, depletion, research reactor, criticality measurements

ID: 205608