A comparison of nuclide production and depletion using MCNPX and ORIGEN-ARP reactor models and a sensitivity study of reactor design parameters using MCNPX for nuclear forensics purposes

A comparison of nuclide production and depletion using MCNPX and ORIGEN-ARP reactor models and a sensitivity study of reactor design parameters using MCNPX for nuclear forensics purposes

The Oak Ridge Isotope Generation and Depletion – Automatic Rapid Proccessing (ORIGEN-ARP) deterministic code has been extensively utilized for determining nuclide concentrations at various specific burnup values for a variety of nuclear reactor designs. Given nuclide concentrations or ratios, such c...

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Bibliographic Details
Main Author: Chambers, Angela Sue
Format: Others
Language:English
Published: 2011
Subjects:
Nuclear forensics
ORIGEN
MCNPX
Burnup values
Depletion calculations
PWR
BWR
CANDU
Sensitivity study
Online Access:http://hdl.handle.net/2152/ETD-UT-2010-05-853

Internet

http://hdl.handle.net/2152/ETD-UT-2010-05-853

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