Steam Oxidation in Accident Conditions

Since the accident at the Fukushima Daiichi Nuclear Plant in March 2011, significant attention has been paid to steam oxidation in a loss of coolant accident (LOCA). Traditionally thought of as an accident scenario driven by a break in a main cooling line, resulting in a loss of pressure in the core and less efficient cooling of fuel rods. Under this condition, steam is generated, and zirconium-based cladding materials experience a number of bulk and microstructural degradation phenomena including but not limited to oxidation, embrittlement, ballooning, blistering, cracking, spallation, and rupture. The behavior of zirconium-based cladding under accident conditions remains a pertinent question, and the development of advanced reactor technologies broadens the impact of steam oxidation to new, hypothetical, accident conditions for both existing and new advanced technology or “accident-tolerant” reactor materials. This article will discuss steam oxidation of conventional water cooled reactor materials, the material responses for a variety of current and proposed reactor materials, and present the current methods for testing materials in simulated accident conditions involving steam.