Zirconium alloy for use in boiling water nuclear reactors - subjected to soln. and ageing heat treatments to improve corrosion resistance

摘要

Zirconium alloy component contg. Sn, fe, Cr, and particles of an intermetallic phase, has improved resistance to accelerated pustular or bubble corrosion in boiling water nuclear reactors. The novelty is that the component microstructure has a segregation of particles with dia. 100-400 angstroms in two-dimensional rows along the grain- and subgrain-boundaries, and distributed throughout the entire component. The alloy pref. contains (a) no Ni, 1.5% Sn, 0.2% Fe, 0.1% Cr, 0.1% O2, rest Zr; or (b) 1.5% Sn, 0.15% Fe, 0.1% Cr, 0.05% Ni, 0.1% O2, rest Zr; and is esp. used for fuel cans or channels contg. a bundle of fuel cans. The component is pref. heated to convert the alpha-phase into the beta-phase and dissolve all intermetallic particles, then quenched to room temp. to avoid pptn. of intermetallic particles; reheating is used to obtain the particles of 100-400 angstroms dia. round the grain boundaries. An esp. pref. treatment is heating 3-60 seconds at 1000-1100 degrees C., water quenching at 800 degrees C/second, then reheating 2-4 hours at 400-600 degrees C. The working life of the components is at least doubled due to the reduced corrosion.