Abstract
Corrosion control in nuclear power systems is examined via addition of LiOH to the coolant at pressurized water reactor (PWR) and supercritical water-cooled reactor (SCWR) operating conditions in stainless steel tubing. The loss of metal to the coolant is analyzed using the voltammetry method. The SEM/EDX analysis of metallographic cross-sections is performed. The results indicate that an adequate pH control is possible for water temperatures up to 500°C. Above this temperature pH control becomes progressively more difficult as dielectric constant and density of water decrease. Significant hydrogen production is detected at 650°C.
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