Engineering formulation of the irradiation growth behavior of zirconium-based alloys for light water reactors

Abstract

Improved Zr-based cladding alloys of which the composition was changed from conventional Zircaloy have been developed by nuclear fuel vendors and utilities. Since the irradiation growth of fuel cladding is one of the most important parameters which determine the dimensional stability of fuel rods and/or fuel assemblies during normal operation, irradiation growth data of such alloys are necessary for safety analysis and proper design of fuel rods and/or fuel assemblies. The irradiation growth behavior of coupon specimens prepared from improved Zr-based alloys for light-water reactor fuel cladding, which have various additive elements and fabrication process, was investigated by conducting an irradiation test at 573 and 593 K under typical PWR coolant conditions up to a fast-neutron fluence of ≈7.8 × 1021 (n/cm2, E >1 MeV) in the Halden reactor in Norway. Based on the dimensional change data measured at interim and final inspections, the amounts of irradiation growth of the improved Zr-based alloys were formulated from the viewpoint of engineering. The trends of the parameters which express the effects of additive elements on irradiation growth behavior were in good agreement with those previously reported, and it was found that the amount of irradiation growth can be expressed by using a summation rule of the effect of each additive element on irradiation growth.