The enhancements and testing for the mcnpx 2.6.0 depletion capability

Nuclear Technology

Volumn 170, Issue 1, 2010, Pages 68-79

  Fensin, Michael L ^a,b   Hendricks, John S ^a   Anghaie, Samim ^a  

^a LOS ALAMOS NATIONAL LABORATORY   (United States)

^b UNIVERSITY OF FLORIDA   (United States)

Author keywords

Burnup; Depletion; MCNPX

Indexed keywords

ACTINIDES; BENCHMARKING; FISSION PRODUCTS; ISOTOPES; REACTION RATES;

BENCHMARK CALCULATIONS; BURN UP; COMPUTATIONAL PERFORMANCE; DEPLETION; MCNPX; MONTE CARLO RADIATION TRANSPORT CODE; THREE DIMENSIONAL GEOMETRY; TRANSPORT CROSS-SECTION;

MONTE CARLO METHODS;

EID: 77950979585     PISSN: 00295450     EISSN: None     Source Type: Journal    
DOI: 10.13182/NT10-2     Document Type: Article

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