The efficiency and operational safety of a pressurized water nuclear power plant are closely linked to the hydraulic conditions in the primary loop. Pressure drops resulting from linear and local losses significantly affect the operation of the reactor coolant pump and consequently the stability of the system. This thesis presents a simplified hydraulic model of the primary loop of a pressurized water reactor, implemented in Python. The analytical model, based on fluid mechanics principles, enables the calculation of pressure drops within individual segments of the system. The model was validated by comparing the results with a similar case from the literature, demonstrating satisfactory agreement. In the final part, simulations of different levels of steam generator tube blockage were performed to analyze the impact of altered hydraulic conditions on the system curve as well as on the operation and loading of the reactor coolant pump. The results showed that tube blockage in the steam generator leads to a significant increase in pressure drops within the system. With increasing blockage, the system characteristics shift, leading to a change in the pump operating point. Increased pressure losses result in greater mechanical loading on the pump, potentially affecting its long-term reliability and efficiency. It was also found that steam generator blockage influences overall heat transfer, reducing core heat removal and lowering steam production on the secondary side.
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