Dynamic simulation of heat transfer in cogenerative nuclear reactors

The developed model which describes the dynamic response of thereactor, primary circuit, and secondary circuit has been analyzed and verified with the relevant models satisfactory.The reactor components and the fuel clad integrity were attained during different step change of reactivity and load demand perturbations.The anticipated scram transients of the desalination units coupled with the nuclear power plant do not affect the reactor safety requirements.The power demand perturbation or steam demand is the most important study to predict the effect of decreasing the power demand on the fuel clad temperature, which maintained below the departure from the nucleate boiling temperature at the reactor applied pressure.• The maximum overshoot in the reactor power in the case of 0.001 positive step reactivity is nearly 7% of the rated value, which do not exceed the safe limits of 20% A figure illustrating the fuel clad temperature response with time at 10, 20, and 25% load demand variations is obtained. This figure indicates that e maximum attained fuel clad temperature not exceeded 340°C.Corresponding to a temperature difference between the fuels clad surface and the coolant temperature; it is fair far enough from the departure from nucleate boiling.• The operating mode curve which describes the secondary side pressure,steam temperature, and steam mass flow rate were obtained in the case of load change.5.3 Suggestions for Future Work• Some components of the pressurized water reactor can be added to the code to study their effect on the system such as pressurizer, circulating pumps, and the condenser.• We can upgrade the present model to deal with different power to water ratios to include the back pressure turbine and extracted steam, enabling the coupling of different desalination techniques to the nuclear power plants.