Preliminary Analysis by means of the TRANSURANUS Code of Mixed Oxide Fuel Rod for Gen IV Lead Fast Reactor

Three fast reactors (FR) are included in the Generation IV (GEN IV) nuclear system concepts: Sodium-cooled Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR) and Lead-cooled Fast Reactor (LFR). FRs have a unique role in the actinides management due to effective fissioning of the transuranic actinides (TRU) recovered from LWRs spent fuel. These GEN IV systems are capable, in principle, to operate with a complete recycle of all the uranium and TRU isotopes, providing a considerable increase in the available fuel resources and a decrease in the long term nuclear waste burden. These innovative systems can operate by using a wide range of fuel types including oxide, carbide, nitride, metal and dispersion fuels with different cladding material options. Concerning fuel technology, most of the gained experience refers to oxide fuels. Cross-cutting and generic open issues concerning fuel and cladding applicable to GEN IV nuclear systems will play an important role in the future nuclear fuel science. In this paper, a preliminary thermo-mechanical study, by means of the TRANSURANUS code (TU), is performed on mixed oxide fuel rods (MOX) operating in a LFR environment. To this purpose the design of the European Lead-cooled SYstem (ELSY), under study in the context of the EURATOM 6th FP, was used as reference. In ELSY, main GEN IV goals are pursued (sustainability, economics, safety and reliability, proliferation resistance and physical protection). The reference fuel rod is similar to that of a classical SFR with MOX, except for cladding material that in ELSY is assumed to be the ferritic-martensitic steel T91. At the next step, the minor actinides (MA) bearing (2-5 wt%) MOX will be considered, while a nitride fuel is the long term option. This preliminary ELSY fuel pin analysis aims at investigating the expected deviations of its performances caused by moving from the reference MOX to a MA-containing mixed oxide fuel relevant for a closed fuel cycle option. Even if the empirical approach of this analysis relies on the carefully tested modelling of TRANSURANUS code, it must be noted that the calculations have been partly performed beyond the validated burnup domain of the code, therefore next due step will be the review of this “blind” introductory analysis to GEN IV LFR fuel rod on the basis of an experimental irradiation in the burnup domain of interest.