A Neutronic Analysis of TRU Recycling in PWRs Loaded with MOX-UE Fuel (MOX with U-235 Enriched U Support)

Youinou, G. ; Bays, S.

Idaho National Laboratory

关键词: Mox;    Uranium;    Recycling;    Void Coefficient Mox;    Coolants;   

DOI  :  10.2172/961912
RP-ID  :  INL/EXT-09-16091
RP-ID  :  DE-AC07-99ID-13727
RP-ID  :  961912

美国|英语

来源: UNT Digital Library

【 摘 要 】

This report presents the results of a study dealing with the homogeneous recycling of either Pu or Pu+Np or Pu+Np+Am or Pu+Np+Am+Cm in PWRs using MOX-UE fuel, i.e. standard MOX fuel with a U235 enriched uranium support instead of the standard tail uranium (0.25%) for standard MOX fuel. This approach allows to multirecycle Pu or TRU (Pu+MA) as long as U235 is available, by keeping the Pu or TRU content in the fuel constant and at a value ensuring a negative moderator void coefficient (i.e. the loss of the coolant brings imperatively the reactor to a subcritical state). Once this value is determined, the U235 enrichment of the MOX-UE fuel is adjusted in order to reach the target burnup (51 GWd/t in this study).