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

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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

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Title: 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
Author: Chambers, Angela Sue
Abstract: 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 calculations can be used in nuclear forensics and nuclear non-proliferation applications to reverse-calculate the type of reactor and specific burnup of the fuel from which the nuclides originated. Recently, Los Alamos National Laboratory has released a version of its probabilistic radiation transport code, MCNPX 2.6.0, which incorporates a fuel burnup feature which can also determine, via the probabilistic Monte Carlo method, nuclide concentrations as a function of fuel burnup. This dissertation compares the concentrations of 46 nuclides significant to nuclear forensics analyses for different reactor types using results from the ORIGEN-ARP and the MCNPX 2.6.0 codes. Three reactor types were chosen: the Westinghouse 17x17 Pressurized Water Reactor (PWR), the GE 8x8-4 Boiling Water Reactor (BWR), and the Canadian Deuterium Uranium, CANDU-37, reactor. Additionally, a sensitivity study of the different reactor parameters within the MCNPX Westinghouse 17x17 PWR model was performed. This study analyzed the different nuclide concentrations resulting from minor perturbations of the following parameters: assembly rod pitch, initial moderator boron concentration, fuel pin cladding thickness, moderator density, and fuel temperature.
Department: Mechanical Engineering
Subject: Nuclear forensics ORIGEN MCNPX Burnup values Depletion calculations PWR BWR CANDU Sensitivity study
URI: http://hdl.handle.net/2152/ETD-UT-2010-05-853
Date: 2010-05

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