Improvement and validation of a sub-channel analysis code for a lead-cooled reactor with wire spacers

The fuel rods in the lead-based fast reactor (LFR) are usually supported by wire spacer, and the presence of the wire results in the complexity of the thermal-hydraulic analysis in rod bundles. In this study, several previous experimental research about lead-bismuth eutectic flowing through the rod bundle are investigated. Based on the collected experimental data, a series of flow and heat transfer correlations in the wire-fixed rod bundle are compared and evaluated. The recommended correlations are then implemented into the in-house sub-channel analysis code SACOS-PB. Furthermore, to account for the changes in the channel cross-section parameters caused by the wire spacer, two optional wire geometric treatments were added to the code: axially-averaged approach and axially-varied approach with extra cross-flow enhancement model. 19-rod bundle and 61-rod bundle tests are used to validate the optimized code. The deviations for the temperature differences of cladding and coolant in internal sub-channels are within 25%, and the axially-varied approach shows a higher prediction accuracy, which proves that the improvement of the SACOS-PB are reliable. This work could provide a reference for the subsequent design and development of the LFR core.