Development and Validation of Subchannel Analysis Program for LMR Based on FEM

Liquid metal reactors (LMRs) involved both sodium fast reactors (SFRs) and lead fast reactors (LFRs) have gained great development over the world due to their nuclear fuel breeding and nuclear waste transmutation capability. For the purpose of meeting the design and construction requirements of LMRs in China, a subchannel program FLARE based on the MOOSE framework for LMRs was developed. The program considers four conservation equations: mass, momentum, and energy in the axial direction, along with momentum in the transverse direction. Additionally, a hybrid approach that integrates the discontinuous Galerkin (DG) and finite element method (FEM) for discretizing conservation equations was employed in FLARE PROGRAM. Advanced closure models such as pressure drop, heat transfer, and material properties of sodium, lead bismuth, and lead were incorporated. In detail, the program integrates the Upgraded Cheng and Todreas (UCTD) wall friction model, Cheng and Todreas (CT) turbulence mixing model, CT transverse heat conduction model, along with some heat transfer correlations with a wide applicability range, while considering geometric effects and flow conditions. To solve the coupled nonlinear partial differential equations, the Jacobians-Free Newton-Krylov (JFNK) algorithm was used. To validate the accuracy of the constitutive models for swirling flow effect implemented in FLARE program, Toshiba 37-pin was computationally simulated, and the results from FLARE PROGRAM were compared with experimental data as well as results from similar programs. The results demonstrate that the computational results from the FLARE program exhibit good agreement with experimental data as well as calculations from similar programs. The maximum error at high and medium flow rates is 12.6% of the inlet-outlet temperature difference, while the error remains within 10% at low flow rates. Moreover, the correctness of the transverse heat conduction correction model was validated through low-flow conditions. To further validate the computational capability of the FLARE program, the KNS 37-pin experiment was selected for simulation. FLARE program exhibits a maximum error of 6.3% in the radial distribution calculation of coolant channel temperature and 8.51% in the axial distribution, demonstrating good overall agreement with experimental values and high computational accuracy. Finally, to validate the FLARE program’s capability for thermal-hydraulic safety analysis of the reactor core, analysis and calculations were performed on the hottest component of the ALFRED conceptual design reactor. Due to the lack of directly comparable experimental data, the results of FLARE program were validated through comparison with those obtained from the similar program SACOS-LMR. The results show that FLARE program provides reasonable calculations for the axial temperature distribution of the hottest fuel rod cladding and core block, and shows good agreement with SACOS-LMR program. This confirms FLARE program’s capability for accurately calculating the temperature distribution of reactor coolant and fuel rods within the core.