Electrochemical Behavior of Accident Tolerant Fuel Cladding Materials under Simulated Light Water Reactor Conditions

After the Fukushima reactor accidents following Japan's March 2011 tsunami, the U.S. Department of Energy engaged with nuclear fuel vendors to develop improved fuels for the current fleet of light water power reactors. General Electric and Oak Ridge National Laboratory have proposed using iron-chrome-aluminum (FeCrAl) ferritic alloys as cladding material for the existing uranium dioxide fuel (UO2). This is a simple approach that leaves unchanged the present coolable geometry in the reactor. FeCrAl alloys have outstanding resistance, in accident conditions, to attack by superheated steam compared to the current zirconium alloys. Since ferritic FeCrAl alloys have not been used before in nuclear power reactors, extensive characterization has been performed to determine their behavior in light water reactor conditions (e.g., normal operation and accident). The present work deals with the electrochemical behavior of the newer alloys in high-temperature water (near 300 degrees C) containing either excess hydrogen gas or excess oxygen. Results show that chromium-containing ferritic FeCrAl have similar electrochemical high-temperature behavior like other common existing reactor alloys containing chromium for passivation (such as X-750, Alloy 600, and Type 304SS). The use of FeCrAl alloy cladding would also eliminate existing common degradation mechanisms such as shadow corrosion in boiling water reactors.