Performance investigation of UN spherical fuel elements in MSR-SF - Haiqing Zhang; Changqing Cao; Peng Wang; Ziwei Li; Xuzhong Kang; Jun Lin (Nuclear Engineering and Technology, 2025)
This study explores Uranium Nitride (UN) fuel performance in Molten Salt Reactors (MSRs), focusing on thermal efficiency, mechanical stability, and fission product retention. UN fuels offer advantages like higher thermal conductivity and heavy metal density compared to traditional uranium oxide (UO2). Key findings include lower kernel temperatures (4.1–20.1 °C reduction) and 25 % longer fuel cycles for UN (85 % TD) versus UO2 (95 % TD). Reduced cladding strain and fission gas release further highlight UN's safety and economic potential. Challenges in MSR designs, such as fuel cycle complexity and material compatibility, are offset by innovations in fuel reprocessing and modular reactor concepts. This work suggests that UN fuel shows potential as a promising candidate for Molten Salt Reactors under modeled conditions, pending experimental validation as a function of burnup.