Published 1991 | Version v1
Book

Investigation of nodular corrosion mechanism for zircaloy products

  • 1. Teledyne-Wah Chang, Albany, OR (United States)

Description

Flat mill products of Zircaloy-4 and controlled-chemistry Zircaloy-2 were processed using different rolling schedules in order to generate different textures and other metallurigical conditions. Tubular products of three heats of controlled-chemistry Zircaloy-2 were also processed using various thermo-mechanical parameters and different tube reduction schedules. Samples were taken for 500 and 520 degrees C steam autoclave tests and texture and microstructure examination. Intermetallic precipitates were analyzed for mean diameters and particle size distributions using optical high magnification photographs. Matrix concentrations of Fe, Cr, and Ni were analyzed by electron microprobe with automated stage control. Results provided in this paper showed that the nodular corrosion resistance for controlled-chemistry Zircaloy-2 was excellent, while that for Zircaloy-4 was rather poor. The Zircaloy-2 tubshell may be annealed up to 663C/2h without detriment to the nodular corrosion resistance of the final size tubing if other fabrication variables are properly controlled. The optimum process for tube reducing is a high Q ratio and high percent of reduction in area. Neither β quench nor α + β treatment after tube extrusion is required to produce cladding tube that is free of nodules in 500 or 520 degrees C autoclave tests. Low basal pole densities are generally related to high corrosion weight gains. The accumulated annealing parameter, ΣA, can generally predict the nodular corrosion behavior of the material, but it cannot account for the higher corrosion weight gain at an intermediate stage and lower weight gain at the final size

Additional details

Publishing Information

Publisher
ASTM.
Imprint Place
Philadelphia, PA (United States)
ISBN
0-8031-1463X
Imprint Title
Zirconium in the nuclear industry
Imprint Pagination
795 p.
Journal Page Range
p. 319-345.

Conference

Title
9. American Society for Testing and Materials (ASTM) international symposium on zirconium in the nuclear industry.
Dates
5-8 Nov 1990.
Place
Kobe (Japan).

Optional Information

Secondary number(s)
CONF-901107--.