skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

Abstract

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within themore » Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

Authors:
 [1];  [1]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
USDOE National Nuclear Security Administration (NNSA)
OSTI Identifier:
1286852
Grant/Contract Number:
AC05-00OR22725
Resource Type:
Journal Article: Accepted Manuscript
Journal Name:
Nuclear Engineering and Design
Additional Journal Information:
Journal Volume: 297; Journal ID: ISSN 0029-5493
Publisher:
Elsevier
Country of Publication:
United States
Language:
English
Subject:
42 ENGINEERING; 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; 98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL PROTECTION

Citation Formats

Mertyurek, Ugur, and Gauld, Ian C. Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs. United States: N. p., 2015. Web. doi:10.1016/j.nucengdes.2015.11.027.
Mertyurek, Ugur, & Gauld, Ian C. Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs. United States. doi:10.1016/j.nucengdes.2015.11.027.
Mertyurek, Ugur, and Gauld, Ian C. 2015. "Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs". United States. doi:10.1016/j.nucengdes.2015.11.027. https://www.osti.gov/servlets/purl/1286852.
@article{osti_1286852,
title = {Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs},
author = {Mertyurek, Ugur and Gauld, Ian C.},
abstractNote = {In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.},
doi = {10.1016/j.nucengdes.2015.11.027},
journal = {Nuclear Engineering and Design},
number = ,
volume = 297,
place = {United States},
year = 2015,
month =
}

Journal Article:
Free Publicly Available Full Text
Publisher's Version of Record

Save / Share:
  • A methodology is described that serves as an alternative to the SAS2H path of the SCALE system to generate cross sections for point-depletion calculations with the ORIGEN-S code. Automatic Rapid Processing (ARP) is an algorithm that allows the generation of cross-section libraries suitable to the ORIGEN-S code by interpolation over pregenerated SAS2H libraries. The interpolations are carried out on the following variables: burnup, enrichment, and water density. The adequacy of the methodology is evaluated by comparing measured and computed spent-fuel isotopic compositions for pressurized water reactor and boiling water reactor systems.
  • The fabrication of mixed uranium-plutonium oxide fuel for the U.K. Fast Breeder Reactor Program is discussed, and the experience gained by the United Kingdom Atomic Energy Authority and British Nuclear Fuels plc over the past 25 yr is used to identify aspects of the manufacturing routes that influence irradiation performance and subsequent recycle. The design of a small-scale production plant for fast reactor mixed-oxide (MOX) fuels is discussed, and its conversion to produce 8 ton/yr (heavy metal) thermal MOX fuel is outlined.
  • Recycling calculations for uranium dioxide (UO/sub 2/) and mixed oxide (MOX) fuel are being carried out in conjunction with Project SR 316 entitled ''Safety Analysis for Thermal Recycling in the Federal Republic of Germany.'' On the basis of the calculated nuclide inventories in spent fuel it is then possible to treat safety issues related to thermal recycling. The reliable determination of fission material quantities and their composition is therefore of particular importance. The OREST program system is used to calculate the nuclide inventories in spent LWR fuel, and a large number of boundary conditions such as reactor and fuel assemblymore » type, burnup and fuel composition can be considered using this system. It makes it possible to carry out the recycling calculations in conjunction with Project SR 316 in a physically satisfactory way. 15 refs., 9 figs., 12 tabs.« less