|About this Abstract
||2020 TMS Annual Meeting & Exhibition
||Accelerated Materials Evaluation for Nuclear Applications Utilizing Irradiation and Integrated Modeling
||Microstructure of HT-9 Cladding After fuel-cladding Chemical Interaction with an Annular U-10Zr Fuel Irradiated to 3.3% FIMA
||Xiang Liu, Luca Capriotti, Tiankai Yao, Jason M Harp, Lingfeng He
|On-Site Speaker (Planned)
U-Zr fuel and ferritic/martensitic HT-9 cladding is the primary fuel system for fast reactors. Although a low smear density (~75%) was often employed to avoid the premature mechanical failure of the cladding, fuel swelling eventually brings the fuel and cladding into contact and leads to fuel-cladding chemical interaction (FCCI). FCCI is a limiting factor that restricts the fuel performance at high burnups. Annual fuel forms are being investigated due to their advantages in back end fuel cycle. Here, we investigated the FCCI layer and nearby cladding regions of an annular U-10Zr fuel with HT-9 cladding irradiated to 3.3% FIMA. Significant amount of fission products diffused into the cladding and formed intergranular precipitates, whereas the outgoing C diffusion led to decarbonization of the cladding. In the FCCI layer, a fine-grained U,Fe-rich phase (wastage), with a high number density of ~15 nm intragranular voids and ~50 nm intergranular voids was found.