| Existing computer codes, developed by Macdonald et al. for modeling water chemistry and estimating the accumulated damage from the stress corrosion cracking in the operating environments of boiling water reactors (BWRs), were superseded by a new code, FOCUS which was improved in this study. This new code predicts water chemistry, electrochemical corrosion potential, and accumulated damage throughout BWR coolant circuits simultaneously under normal water chemistry or hydrogen water chemistry.Using the frame of FOCUS, a new simulation code for the Pressurized Water Reactor (PWR) was developed in this study. In order to simulate the PWR, different radiological and chemical models for the calculation of individual chemical species concentrations were developed and equipped in this new code for PWRs. This new code was named P-ECP. Also, the pH model capable of calculating pH values with variations of temperatures and concentrations of the boric acid and lithium hydroxide was developed.The surface area of the steam generator tubes accounts for seventy-five percent of the pressure boundary in typical Westinghouse three-loop PWRs therefore, including the stream generator tubes into the corrosion simulation is critical in the simulation of the corrosion characteristics of PWRs. High temperature and high pressure experimental work was carried out in this study in order to seek the electrokinetic parameters of Alloys 600 and 690. Oxide film properties formed on the surface of the Alloy 690 were investigated by the Mott-Schottky analysis and the Point Defect Model. The obtained electrokinetic parameters were incorporated in the P-ECP code. The simulation results using the new code were in well agreement with the experimental work performed by other groups.The newly developed P-ECP code is capable of calculating pH values and conductivity considering the radiolitic effect under the presence of the boric acid and lithium hydroxide from room temperature to the operating temperature of PWRs. In addition, ECP values throughout the reactor coolant circuits of Type 304 stainless steel and Alloys 600 and 690 can be simulated with the various reactor coolant environments and temperatures under the presence of the hydrogen and oxygen in the coolant from the initial heat-up to normal power operation. |
