MOX Cross-Section Libraries for ORIGEN-ARP

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Description

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 ... continued below

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Gauld, I.C. July 1, 2003.

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Description

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.

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  • Report No.: ORNL/TM-2003/2
  • Grant Number: DE-AC05-00OR22725
  • DOI: 10.2172/885544 | External Link
  • Office of Scientific & Technical Information Report Number: 885544
  • Archival Resource Key: ark:/67531/metadc891630

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Office of Scientific & Technical Information Technical Reports

Reports, articles and other documents harvested from the Office of Scientific and Technical Information.

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  • July 1, 2003

Added to The UNT Digital Library

  • Sept. 23, 2016, 2:42 p.m.

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  • Dec. 7, 2016, 5:55 p.m.

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Gauld, I.C. MOX Cross-Section Libraries for ORIGEN-ARP, report, July 1, 2003; [Tennessee]. (digital.library.unt.edu/ark:/67531/metadc891630/: accessed December 11, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.