An improved core thermal-hydraulic model for coastdown transient in pressurized water reactor

Abstract

The safety of the nuclear reactor revolves around the accurate analysis of rapid flow transient for the design and manufacturing of reactor coolant pumps. In this article, the coastdown transient initiated by the loss of offsite power is simulated. In this case, the pumps are operated by the inertia of the flywheel, therefore, the reliable operation of reactor coolant pumps is the key to the safety of the nuclear reactor. A new hydraulic, as well as the thermal model, is developed for simulating various core parameters during the coastdown period. The present hydraulic model accounts for both the pump half-time and the loop half-time, which is used to increase the accuracy of predicted results over a longer period of time. The results predicted by the hydraulic model are incorporated into a thermal model, which also includes the decay heat following the reactor shutdown. This new model depends upon the core time constant, loop time constant, pump half-time, and hydraulic constant coefficient. The predicted results of flow rate, pressure, temperature, and departure from nucleate boiling ratio are compared with the experimental data and have found good agreement between the two cases. Finally, the departure from nucleate boiling ratio shows that the transient behavior of the reactor is moving toward safety.