Thermal-hydraulic analysis of flow blockage in a supercritical water-cooled fuel bundle with sub-channel code

Abstract Sub-channel code is nowadays the most applied method for safety analysis and thermal-hydraulic simulation of fuel assembly. It plays an indispensable role to predict the detail thermal-hydraulic behavior of the supercritical water-cooled reactor (SCWR) fuel assembly because of the strong non-uniformity within the fuel bundle. Since the coolant shows a strong variation of physical thermal property near the pseudo critical line, the local blockage in an assembly of a SCWR is of importance to safety analysis. Due to the low specific heat of supercritical water with high temperature, the blockage and the subsequent flow reduction at the downstream of the blockage will yield particular high cladding temperature. To analyze the local thermal-hydraulic parameters in the supercritical water reactor-fuel qualification test (SCWR-FQT) fuel bundle with a flow blockage caused by detachment of the wire wrap, the sub-channel code COBRA-SC is unitized. The code is validated by some blockage experiments, and it reveals a good feasibility and accuracy for the SCWR and blockage flow analysis. Some new models, e.g. the axial and circumferential heat conduction model, turbulent mixing models, pressure friction models and heat transfer correlations, are incorporated in COBRA-SC code. And their influence on the cladding temperature and mass flow distribution are evaluated and discussed. Based on the results, the appropriate models for description of the flow blockage phenomenon in SCWR assembly is identified and recommended. A transient analysis of the SCWR-FQT bundle with blockage flow is also presented and discussed by COBRA-SC code. The results indicate that the peak cladding temperature caused by blockage in the fuel assembly can be reduced effectively by the safety measures of SCWR-FQT.

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