Corrosion behavior of stainless steel 316 in sub- and supercritical aqueous environments: Effect of LiOH additions

• Corrosion control via addition of LiOH in stainless steel reactor is tested.
• Subcritical and supercritical water flow conditions are examined.
• At the highest temperature of 650 °C no inhibiting effect of LiOH is observed.
• Hydrogen generation is detected in the reactor at supercritical conditions.

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.