Irradiation Growth Behavior of Improved Alloys for Fuel Cladding

Abstract

New Zr alloys for fuel cladding with different compositions from conventional ones have been de-veloped to increase the safety of nuclear power plants and to utilize existing nuclear power plants more effectively. Since the irradiation growth of fuel cladding is one of the most important parameters regarding the dimensional stability of a fuel rod and / or fuel assembly during irradiation, the irradiation growth behavior of the improved Zr alloys for light-water reactor fuel cladding was investigated. Coupon specimens, which were prepared from fuel cladding tubes with improved Zr alloys, were irra-diated in the Halden reactor in Norway at temperatures of 300 and 320 ℃ under a typical water chem-istry condition of a PWR and at 240 ℃ under the coolant condition of the Halden reactor up to a fast neutron fluence of ～ 8 × 10 25 （ 1 / m 2 , E ＞ 1 MeV ） . During and after the irradiation test, the amount of irradiation growth of each specimen was evaluated. The effect of the difference in alloy composition on irradiation growth behavior seemed insignificant if the other conditions, such as the final heat treat ment condition at fabrication, the irradiation temperature and the amount of hydrogen precharged in the specimen, were the same.