Creep of zirconium alloys - report on irradiation experiment carried out in FBTR

Alloys of Zirconium (Zircaloy-2 and Zr-2.5%Nb alloy) are used as pressure tube materials in Pressurised Heavy Water Reactors (PHWRs ) due to properties such as low thermal neutron absorption cross-section, good corrosion resistance in hot pressurised water and adequate mechanical strength at the operating temperatures. Pressure tubes are critical components of PHWRs containing fuel bundles and coolant at high pressure. The creep rate of zirconium alloys increases substantially in the presence of radiation. The data on the in-reactor creep rate of zirconium alloys, developed indigenously, during the actual service condition is required to predict the performance of the pressure tubes during their service lifetime. An irradiation experiment was designed and carried out in Fast Breeder Test Reactor (FBTR) to determine the in-reactor creep behaviour of these alloys (Zircaloy-2 and Zr-2.5%Nb alloy). Irradiation of zirconium alloy specimens in FBTR offers advantage of attaining higher damage rates in a shorter time as compared to irradiation in PHWRs due to the high flux and the hard neutron spectrum of FBTR. Miniature pressure tubes with 15.3 mm O.D. and wall thickness of 0.65 mm were used in the form of pressurised capsules in this experiment. The pressurised capsules were subjected to biaxial stress conditions with stress levels that are allowable in pressure tubes. Irradiation of the pressurised capsules was carried out in FBTR for different durations and the creep strain accumulated was measured in the hot cells. The in-reactor creep rates for indigenous Zircaloy-2 and Zr-2.5%Nb alloy are comparable to those reported in international open literature, which gives confidence in the performance of the indigenously developed alloys. (author)