Steam Oxidation in Accident Conditions

Elizabeth Sooby Wood, Jordan Vandegrift, Brian Jaques

Research output: Chapter in Book/Report/Conference proceedingChapter

Abstract

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.

Original languageAmerican English
Title of host publicationReference Module in Materials Science and Materials Engineering
StatePublished - 1 Jan 2020

Keywords

  • nuclear cladding
  • nuclear fuel
  • oxidation kinetics
  • steam oxidation

EGS Disciplines

  • Materials Science and Engineering

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