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• Behavior of Cr-coated clad under integral LOCA test is simulated in FEA. • Current burst criteria for Zircaloy is ineffective to capture Cr-coated concept. • New overstress and plasticity instability criteria for the Cr coating are proposed. • Coating thickness is found to be numerically insensitive to burst temperature. Chromium (Cr) coated cladding is one of the most promising accident-tolerant fuels for pressurized water reactors. The behavior of Cr-coated cladding under the accident conditions is crucial to quantify its implication on safety, licensing and economics. This work focuses on the modeling of the burst behavior of Cr-coated cladding under Loss of Coolant Accidents (LOCAs). For this purpose, the material properties and behavior models of Zircaloy and chromium have been studied and implemented into the ABAQUS. Then the ABAQUS models are compared against several burst tests done by Korea Atomic Energy Research Institute (KAERI) and Électricité de France (EDF). During the comparison process, different high temperature creep models, zircaloy cladding burst criteria and coating thicknesses are adopted to investigate the thermo-mechanical performance and burst behavior of Cr-coated claddings under LOCA. Based on the comparative computational analysis, 100% and 85% Cr fracture strength and the strain rate of 0.23 /s and 0.07 /s are proposed as a new overstress burst criterion for coated cladding under a heating rate from 5 K/s and 10 K/s. The effect of coating thickness on the proposed criteria and the ballooning behavior of the coated cladding was predicted to be insignificant. The proposed criteria are beneficial for better understanding of the coated clad burst phenomena and the best estimation of coated clad performance under LOCA scenarios. [ABSTRACT FROM AUTHOR]

• Equilibrium equations and compatibility equations were conducted respectively to consider coupling effects of subassemblies in SFR. • Calculation for row bowing was performed with the aid of FADAC code in the present study. • Row bowing was analyzed in detail under different geometric parameters for wrapper tube, even core baffle was taken into consideration. • Reaction forces and deflection parameters were obtained by calculation, which are meaningful for safety evaluation and core structural design. Subassemblies in sodium-cooled fast reactor are different from pressurized water reactor in both geometry and assembly arrangement. In sodium-cooled fast reactor, assembly thermal deformation and narrow clearance between adjacent assemblies would cause them contact with each other, and reaction force between neighboring pads would occur. Under that circumstance, assembly deformation would be decided by both thermal load and mechanical load, which means significant coupling effect would lead to more complicated deformation behaviors. Therefore, so as to deeply study such thermal–mechanical coupling effect in reactor core, it is meaningful to conduct theoretical analysis and numerical simulation on such behaviors, which is indispensable for reactor safety analysis and structural design. In the present study, equilibrium equations and compatibility equations were conducted respectively for acquiring deformation parameters of assemblies in the row, which were already implemented into FADAC code and can perform calculation for row bowing. Furthermore, reaction forces and assembly deflections were analyzed in detail. Besides, sensitivity analysis were performed for row bowing calculation, in which wrapper tube thickness, distance across flat for wrapper tube, pad clearance and effects of core baffle were all taken into consideration. The present theoretical research for assembly deformation in SFR is essential for reactor safety and structural design, and will lay a solid foundation for assembly deformation analysis in the whole core. [ABSTRACT FROM AUTHOR]