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

軽水炉用ジルコニウム合金の照射成長挙動の定式化

垣内 一雄 ; 天谷 政樹  ; 宇田川 豊  

Kakiuchi, Kazuo; Amaya, Masaki; Udagawa, Yutaka

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 conditions, was investigated by conducting an irradiation test at 573 and 593 K under typical PWR coolant conditions up to a fast-neutron fluence of 7.810 (n/cm , 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.