The simulation of fuel composition requires coupled calculations of neutron transport and burnup. It is generally assumed that the neutron flux density and cross-sections remain constant within a burnup step. However, when there are strong absorber poisons present, the reaction rates of the absorbers change too rapidly over time, necessitating extremely fine step sizes to ensure computational accuracy, which in turn leads to low computational efficiency. As a type of accident tolerant fuel (ATF), fully ceramic micro-encapsulated (FCM) fuel is a promising new type of nuclear fuel. Accelerated algorithms for burnup calculations of FCM fuel containing gadolinium isotopes have been developed based on the ALPHA code, including the projected predictor–corrector (PPC), the log-linear rate (LLR), and the high-order predictor–corrector (HOPC) methods (including CE/LI, CE/QI, LE/LI, and LE/QI). The performances of different algorithms under the two forms of Gd₂O₃ existence were analyzed. The numerical results show that the LE/QI method performs the best overall. For Gd₂O₃ existing in both forms, the LE/QI algorithm can maintain accuracy with a burnup step size of up to 1.0 GWd/tU, keeping the infinite multiplication factor k_inf within 100 pcm, and it exhibits high accuracy in simulating the atomic number densities of Gd-155 and Gd-157 throughout the burnup process.

As a promising nuclear technology, molten salt reactors (MSRs) have a bright future in the energy sector due to their unique advantages such as high efficiency, safety, and fuel flexibility. However, the accurate analysis of fission products in MSRs requires reliable fission yield data. Current reactor burnup analysis often uses the ORIGEN2 code, whose fission yield libraries mainly originate from the outdated 1970s ENDF/B-VI nuclear database, thus risking data obsolescence. This study evaluates ORIGEN2’s fission yield libraries (THERMAL, PWRU, PWRU50) against the modern ENDF/B-VIII.0 library. Through a comprehensive comparative analysis of Oak Ridge National Laboratory’s Molten Salt Reactor Experiment (MSRE) model, numerical simulations reveal library-dependent differences in MSR burnup characteristics. The PWRU library best matches ENDF/B-VIII.0 for U-235-fueled cases in keff results, while the PWRU50 library has minimal keff deviation in U-233-fueled setups. Moreover, in both fuel cases, the fission yield library was found to significantly affect the activity of key radionuclides, including Kr-85, Kr-85m, I-133m, Cs-136, Sn-123, Sn-125, Sn-127, Sb-124, Sb-125, Cd-115m, Te-125m, Te-129m, etc. Additionally, the fission gas decay heat power calculated via the ORIGEN2 library is over 20% lower than that from the ENDF/B-VIII.0 library tens of days after shutdown, mainly due to differences in long-lived Kr-85 production. These findings highlight the need to update traditional fission yield libraries in burnup codes. For next-generation MSR designs, this is crucial to ensure accurate safety assessments and the effective development of this promising energy technology.