Current Perspectives on Zirconium Use in Light Water Reactor Fuel and Its Continued Use in Nuclear Power

Abstract

Since the introduction of Zircaloys as light water reactor (LWR) nuclear fuel cladding materials about 60 years ago, significant irradiation experience has been accumulated on zirconium alloys. The need for interim dry storage of spent nuclear fuel was also not known then. As a result, the requirements determining the use of zirconium alloys for LWR fuel cladding have increased significantly beyond the initial focus on waterside corrosion resistance, oxygen pickup, and oxygen embrittlement. These additional evolving factors include hydrogen pickup, hydride embrittlement, hydride reorientation during dry storage, microstructural stability under neutron irradiation, irradiation-induced dimensional changes (growth and creep), the amount of hydrogen generated under accident conditions, evaluation of loss-of-coolant accident embrittlement due to oxygen and hydrogen, evaluation of loss-of-coolant accident breakaway oxidation, and evaluation of response to the reactivity insertion accidents. The impact of these evolving factors on the current zirconium alloys (composition, processing, and microstructure) are discussed in this paper. Summaries of selected past contributions of the author related to zirconium alloy material understanding for improvement of their LWR performance are presented. Because the continued use of zirconium alloys in the nuclear industry is related to the survival and expansion of nuclear power, factors critical for the expansion of nuclear power are also discussed for the current situation in the United States. Corrective actions taken by the nuclear industry to mitigate the consequences of the three major accidents in power reactors are discussed, as are ways to maintain the use of zirconium in nuclear power in the future.