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Title: Monte-Carlo Code (MCNP) Modeling of the Advanced Test Reactor Applicable to the Mixed Oxide (MOX) Test Irradiation

Abstract

Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, and 40 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Disposition Program (FMDP). The fuel burnup analyses presented in this study were performed using MCWO, a welldeveloped tool that couples the Monte Carlo transport code MCNP with the isotope depletion and buildup code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements.

Authors:
;
Publication Date:
Research Org.:
Idaho National Laboratory (INL)
Sponsoring Org.:
DOE - NE
OSTI Identifier:
911248
Report Number(s):
INL/EXT-05-00599
TRN: US0704491
DOE Contract Number:  
DE-AC07-99ID-13727
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
73 - NUCLEAR PHYSICS AND RADIATION PHYSICS; ACTINIDES; BUILDUP; BURNUP; FISSILE MATERIALS; IRRADIATION; LANL; NEUTRON MONITORS; ORNL; OXIDES; PLUTONIUM; SIMULATION; TEST REACTORS; TRANSPORT; Advanced Test Reactor; Irradiation; MCNP; MOX

Citation Formats

Chang, G S, and Pederson, R C. Monte-Carlo Code (MCNP) Modeling of the Advanced Test Reactor Applicable to the Mixed Oxide (MOX) Test Irradiation. United States: N. p., 2005. Web. doi:10.2172/911248.
Chang, G S, & Pederson, R C. Monte-Carlo Code (MCNP) Modeling of the Advanced Test Reactor Applicable to the Mixed Oxide (MOX) Test Irradiation. United States. https://doi.org/10.2172/911248
Chang, G S, and Pederson, R C. Fri . "Monte-Carlo Code (MCNP) Modeling of the Advanced Test Reactor Applicable to the Mixed Oxide (MOX) Test Irradiation". United States. https://doi.org/10.2172/911248. https://www.osti.gov/servlets/purl/911248.
@article{osti_911248,
title = {Monte-Carlo Code (MCNP) Modeling of the Advanced Test Reactor Applicable to the Mixed Oxide (MOX) Test Irradiation},
author = {Chang, G S and Pederson, R C},
abstractNote = {Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, and 40 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Disposition Program (FMDP). The fuel burnup analyses presented in this study were performed using MCWO, a welldeveloped tool that couples the Monte Carlo transport code MCNP with the isotope depletion and buildup code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements.},
doi = {10.2172/911248},
url = {https://www.osti.gov/biblio/911248}, journal = {},
number = ,
volume = ,
place = {United States},
year = {2005},
month = {7}
}