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Abstract
Recent efforts to develop high-fidelity, multi-physics safety analysis methodologies aim to assess realistic safety margins. However, they remain conservative using initial conditions based on the hot rod with maximum power for the whole reactor core. This study presents the development and verification of a coupled code for pin-wise reactor core analysis using the subchannel analysis code CUPID and the fuel performance code GIFT. The goal was to generate accurate pin-wise fuel rod conditions during normal operation for safety analysis. CUPID was applied for thermal–hydraulic analysis considering fuel behavior, and GIFT was extended to multiple fuel rods. The coupled code was verified for a single assembly to examine subchannel geometry deformation, followed by a practical simulation for a quarter reactor core during the first cycle. Key parameters for safety analysis were examined, revealing that maximum stored energy occurred in a specific fuel rod rather than the hot rod with maximum power.
Published Version
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