PWR ENDF/B-VII cross-section libraries for ORIGEN-ARP

Conference · Sun Jul 01 04:00:00 EDT 2012

OSTI ID:22105754

McGraw, C. ^[1]; Ilas, G. ^[2]

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1. Dept. of Nuclear Engineering, Texas A and M Univ., 3133 TAMU, College Station, TX 77843-3133 (United States)
2. Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6172 (United States)

New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% {sup 235}U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time. (authors)

• Research Organization: American Nuclear Society, Inc., 555 N. Kensington Avenue, La Grange Park, Illinois 60526 (United States)
• OSTI ID: 22105754
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BURNUP
CALCULATION METHODS
CROSS SECTIONS
DESIGN
NUCLEAR DATA COLLECTIONS
PWR TYPE REACTORS
RADIOCHEMICAL ANALYSIS
SIMULATION
SLIGHTLY ENRICHED URANIUM
SPENT FUELS
TWO-DIMENSIONAL CALCULATIONS
URANIUM 235
VALIDATION