MOX Cross-Section Libraries for ORIGEN-ARP
Abstract
The use of mixed-oxide (MOX) fuel in commercial nuclear power reactors operated in Europe has expanded rapidly over the past decade. The predicted characteristics of MOX fuel such as the nuclide inventories, thermal power from decay heat, and radiation sources are required for design and safety evaluations, and can provide valuable information for non-destructive safeguards verification activities. This report describes the development of computational methods and cross-section libraries suitable for the analysis of irradiated MOX fuel with the widely-used and recognized ORIGEN-ARP isotope generation and depletion code of the SCALE (Standardized Computer Analyses for Licensing Evaluation) code system. The MOX libraries are designed to be used with the Automatic Rapid Processing (ARP) module of SCALE that interpolates appropriate values of the cross sections from a database of parameterized cross-section libraries to create a problem-dependent library for the burnup analysis. The methods in ORIGEN-ARP, originally designed for uranium-based fuels only, have been significantly upgraded to handle the larger number of interpolation parameters associated with MOX fuels. The new methods have been incorporated in a new version of the ARP code that can generate libraries for low-enriched uranium (LEU) and MOX fuel types. The MOX data libraries and interpolation algorithms in ORIGEN-ARPmore »
- Authors:
- Publication Date:
- Research Org.:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Sponsoring Org.:
- USDOE
- OSTI Identifier:
- 885544
- Report Number(s):
- ORNL/TM-2003/2
TRN: US0604048
- DOE Contract Number:
- DE-AC05-00OR22725
- Resource Type:
- Technical Report
- Country of Publication:
- United States
- Language:
- English
- Subject:
- 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; 29 ENERGY PLANNING, POLICY AND ECONOMY; ALGORITHMS; BENCHMARKS; BURNUP; COMPUTERS; CROSS SECTIONS; FUEL ASSEMBLIES; INTERPOLATION; INVENTORIES; ISOTOPES; LICENSING; NUCLEAR POWER; OECD; RADIATION SOURCES; SAFEGUARDS; SAFETY; URANIUM; VERIFICATION
Citation Formats
Gauld, I C. MOX Cross-Section Libraries for ORIGEN-ARP. United States: N. p., 2003.
Web. doi:10.2172/885544.
Gauld, I C. MOX Cross-Section Libraries for ORIGEN-ARP. United States. https://doi.org/10.2172/885544
Gauld, I C. 2003.
"MOX Cross-Section Libraries for ORIGEN-ARP". United States. https://doi.org/10.2172/885544. https://www.osti.gov/servlets/purl/885544.
@article{osti_885544,
title = {MOX Cross-Section Libraries for ORIGEN-ARP},
author = {Gauld, I C},
abstractNote = {The use of mixed-oxide (MOX) fuel in commercial nuclear power reactors operated in Europe has expanded rapidly over the past decade. The predicted characteristics of MOX fuel such as the nuclide inventories, thermal power from decay heat, and radiation sources are required for design and safety evaluations, and can provide valuable information for non-destructive safeguards verification activities. This report describes the development of computational methods and cross-section libraries suitable for the analysis of irradiated MOX fuel with the widely-used and recognized ORIGEN-ARP isotope generation and depletion code of the SCALE (Standardized Computer Analyses for Licensing Evaluation) code system. The MOX libraries are designed to be used with the Automatic Rapid Processing (ARP) module of SCALE that interpolates appropriate values of the cross sections from a database of parameterized cross-section libraries to create a problem-dependent library for the burnup analysis. The methods in ORIGEN-ARP, originally designed for uranium-based fuels only, have been significantly upgraded to handle the larger number of interpolation parameters associated with MOX fuels. The new methods have been incorporated in a new version of the ARP code that can generate libraries for low-enriched uranium (LEU) and MOX fuel types. The MOX data libraries and interpolation algorithms in ORIGEN-ARP have been verified using a database of declared isotopic concentrations for 1042 European MOX fuel assemblies. The methods and data are validated using a numerical MOX fuel benchmark established by the Organization for Economic Cooperation and Development (OECD) Working Group on burnup credit and nuclide assay measurements for irradiated MOX fuel performed as part of the Belgonucleaire ARIANE International Program.},
doi = {10.2172/885544},
url = {https://www.osti.gov/biblio/885544},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Tue Jul 01 00:00:00 EDT 2003},
month = {Tue Jul 01 00:00:00 EDT 2003}
}