PRELIMINARY MODELLING OF U-PU MOX FUEL IN CANDU GEOMETRIES USING BISON
13th International Conference on CANDU Fuel - 2016 Aug. 15-18


Presented at:
13th International Conference on CANDU Fuel
2016 Aug. 15-18
Location:
Kingston, ON Canada
Session Title:
Session 4: Advanced Fuel Cycles - MOX

Authors:
A. Prudil (Canadian Nuclear Laboratories)
M. Welland (Canadian Nuclear Laboratories)
W. Richmond (Canadian Nuclear Laboratories)
S. Yatabe (CNL)
  

Abstract

Preliminary models of Uranium-Plutonium Mixed Oxide (MOX) fuel elements from two experimental bundles irradiated in the National Research Universal (NRU) have been developed in the BISON fuel performance code. The code couples heat transport, thermomechanics, temperature and burnup-dependent material properties, fission product behaviour in the fuel, and cladding deformation within a Finite Element framework. Model predictions compare favourably with post-irradiation measurements of cladding deformation and fission gas release. The fuel was composed of 3.1% 5.3% Pu in depleted uranium and obtained peak linear powers of 41 kW m-1 and burnups of 375 MWh (kgHE)-1. The samples represent two methods of MOX fabrication, which result in similar Pu homogenization in elements based on the 37-element bundle design.

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