SYMPOSIA PAPER Published: 01 February 2018
STP159720160066

Effects of Chemistry and Microstructure on Corrosion Performance of Zircaloy-2-Based BWR Cladding

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In boiling water reactors (BWRs), Zircaloy-2 is typically the material of choice for fuel cladding. With increased focus on fuel performance and reliability, consistent cladding corrosion performance under varying BWR conditions has become an acknowledged requirement for modern BWR fuel. To support optimization of corrosion performance, modifications to alloy chemistry and manufacture conditions can be made, supported by ex-reactor corrosion testing and demonstrated through in-reactor operation. In this work, the combined effects of alloy chemistry and fabrication process were investigated. The effect of alloy chemistry is often complicated by variation within a production ingot. In this work, additional measurements were intentionally taken at the tube-shell stage (before cold pilgering) so that chemistry more representative of the final cladding could be obtained and the effect of iron, nickel, and tin could be examined more closely. Although chemistry range of Zircaloy-2 was the primary focus, the effect of iron above the ASTM limit in a new alloy was also explored. Three types of cladding fabrication processes were considered. Two types included a heat treatment applied after the first cold pilgering and annealing, and the treatment was limited to the outer portion of the cladding and differed in the temperature of the heat treatment; the third type had no special heat treatments after the first cold pilgering and annealing. The weight gain of the three types of cladding with varying compositions under two corrosion test conditions (400°C steam and 410/520°C two-stage steam) are discussed in terms of the initial second-phase particle sizes obtained from transmission electron microscopy examinations. Based on poolside inspections and hot-cell examinations, the in-reactor corrosion performance of the two cladding types up to typical discharge exposures are discussed in the terms of corrosion margin, microstructure evolution, and correlation with ex-reactor corrosion test data.

Author Information

Lin, Yang-Pi
Global Nuclear Fuel—Americas, Wilmington, NC
White, David, W.
Global Nuclear Fuel—Americas, Wilmington, NC
Lutz, Dan, R.
Global Nuclear Fuel--Americas, Sunol, CA
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Details
Developed by Committee: B10
Pages: 881–908
DOI: 10.1520/STP159720160066
ISBN-EB: 978-0-8031-7642-3
ISBN-13: 978-0-8031-7641-6