Abstract
Although the primary side metal alloy of Pressurised Water Reactor (PWR) has high corrosion resistance and low overall corrosion rate, corrosion still occurs during the normal operation. Corrosion products will be released into the coolant after the metal materials in primary side are corroded. When these corrosion products flow through the core, they may be deposited on the surface the cladding as fuel crud. If the core boils, the corrosion products will be enriched in these areas and form thick crud. In new edition of HAD102/07-2020” Nuclear Power Plant Reactor Core Design”, it is clearly required that the design analysis should consider the deterioration of fuel rod heat transfer on cladding surface caused by the deposition of corrosion products generated during normal operation of reactor coolant system. Therefore, it is necessary to analyze the influence of fuel crud on the heat transfer performance of fuel rods under accident conditions, especially the typical accident conditions with fuel pellet center temperature and cladding temperature as acceptance criteria, such as Reactivity Initiated Accident (RIA). In order to evaluate the effect of fuel crud on RIA, the mechanism of crud formation in PWR nuclear power plant was investigated in this analysis, and the equivalent thermal conductivity model considering crud was developed independently. The fuel rod transient code BIRCH was used to calculate and analyze the RCCA Ejection Accident (REA) and the Uncontrolled Single RCCA Withdrawal at Power. The results show that the peak fuel pellet temperature and peak cladding temperature are obviously increased after crud is considered, but they still meet the requirements of accident acceptance criteria.