Physics-informed Coarse-mesh Method and Full-core Refined Thermohydraulic Analysis

Traditional safety analysis methods and computational fluid dynamics (CFD) methods fall short of meeting the increasing demands for refined transient safety analysis of nuclear reactors, as system-level codes lack the spatial resolution to capture local thermohydraulic phenomena while high-fidelity CFD remains computationally prohibitive for full-core transient simulations, particularly for long-duration accident scenarios and multi-physics applications. This paper therefore introduces a physics-informed coarse-mesh method (PICM) as an alternative approach that aims to balance computational efficiency with predictive accuracy, offering a new scale option that bridges the gap between conventional subchannel analysis and detailed CFD simulations. The PICM was developed using a coarse-mesh discretization strategy, and instead of resolving fine-scale flow details, the PICM relied on flow regime maps and empirical closure models to correct wall and interphase interaction source terms, with dedicated models introduced to account for the effects of internal structures such as wire wraps and mixing vane spacers, implicitly considering the resulting crossflow mixing, additional pressure drop, and heat transfer enhancement. The PICM was rigorously validated against a series of experimental benchmark datasets covering both single-phase and two-phase conditions. For single-phase flow, validation was performed against smooth rod bundle experiments and the PSBT 5×5 rod bundle mixing experiments, confirming the method’s capability to predict velocity distributions and spacer-induced cross-flow mixing. For two-phase flow, validation was conducted against the PSBT 5×5 rod bundle steady-state and transient boiling experiments, including power ramp and flow reduction scenarios, demonstrating accurate prediction of void fraction distributions and boiling heat transfer. Systematic comparisons were performed against conventional CFD results to assess predictive performance and computational cost. Validation results show excellent agreement between PICM predictions and experimental measurements under both single-phase and two-phase conditions, with key thermohydraulic phenomena including inter-subchannel crossflow, void fraction distribution, velocity fields, and temperature fields effectively captured with high-fidelity. Compared with CFD method, the PICM achieves a computational speedup of approximately three orders of magnitude while maintaining comparable accuracy in global integral parameters and most local quantities of engineering interest, making full-core transient analyses feasible within practical engineering timescales. For the pressurized water reactor (PWR) core equipped with mixing vane spacers, the PICM successfully captures spacer-induced crossflow and the associated heat transfer enhancement. A rod ejection accident transient analysis was further performed, assuming that under steady-state full-power operating conditions, one control rod assembly was ejected, introducing a strong positive reactivity into the adjacent core region. The computational results reveal that the positive reactivity insertion leads to a pronounced local elevation in coolant temperature and void fraction, demonstrating the PICM’s capability to capture rapid transient responses and localized safety margins. For the wire-wrapped lead-bismuth fast reactor core, the results accurately reflect the wire-induced periodic swirling flow and the resulting circumferential non-uniform heat transfer. In addition, a multi-physics coupling analysis of oxidation, corrosion, and deposition in lead-bismuth fast reactors was conducted, demonstrating the PICM’s extensibility beyond pure thermal hydraulics to address material degradation and coolant chemistry effects. The PICM demonstrates excellent geometric adaptability across different fuel assembly designs, including helical fuel and annular fuel, and is suitable for a wide range of advanced reactor concepts. With its demonstrated efficiency, accuracy, and extensibility, the PICM provides a valuable and practical tool for high-fidelity, full-core transient safety assessments, supporting refined reactor design optimization and comprehensive multi-physics coupling analyses in next-generation nuclear systems.