Automatic Whole Core Depletion & Criticality Calculations by MCNPX 2.7.0

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Abstract

Different approaches to perform automatic whole core criticality & depletion calculations in a research reactor using MCNPX 2.7.0 are presented. An approximate method is to use the existing symmetries of the burned fuel material distribution in the core, i.e., the axial, radial and azimuth symmetries around the core center, in order to significantly reduce the computation time. In this case it is not necessary to give a unique material number to each burn up cell. Cells having similar burn up and power, achieved during similar irradiation history at same initial fuel composition, will experience similar composition evolution and can therefore be given the same material number. To study the impact of the number of unique burn up materials on the computation time and utilized RAM memory, several MCNPX models have been developed. The paper discusses the accuracy of the model on comparison with measurements of BR2 operation cycles in function of the number of unique burn up materials and the impact of the used Q-value (MeV/fission) of the recoverable fission energy.

Details

Original languageEnglish
Title of host publicationPhysor 2012
Place of PublicationLaGrange Park, IL, United States
Publication statusPublished - 15 May 2012
Event2012 - PHYSOR: Advances in Reactor Physics - Knoxville Convention Center, Knoxville, United States
Duration: 15 Apr 201220 Apr 2012
http://meetingsandconferences.com/physor2012/

Conference

Conference2012 - PHYSOR
CountryUnited States
CityKnoxville
Period2012-04-152012-04-20
Internet address

Keywords

  • MCNPX 2.7.0, whole core depletion/criticality analysis, unique burn up material.

ID: 342692