Irradiation Testing of Zirconium Alloys and Stainless Steel in Fast Breeder Test Reactor, India

摘要

Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, India is a sodium cooled fast reactor with a maximum neutron flux of about 3 × 10~(15) n/cm~2/s. The neutron spectrum in the FBTR is hard and the damage rate attained in structural specimens is high. Compact pressurized capsules of zirconium alloys have been developed and subjected to irradiation in the FBTR to a fluence level of 1.1 × 10~(21) n/cm~2 (E ＞1 MeV) at temperatures of 306 to 319°C to determine the irradiation creep rate of indigenously developed zirconium alloys (Zircaloy-2 and Zr-2.5%Nb alloy). To assess the changes in mechanical properties of FBTR grid plate material (modified type 316 stainless steel) due to prolonged low dose exposure, an accelerated irradiation test with dose levels up to 2.6 dpa (at 350°C) has also been carried out using miniature tensile test and disk specimens. Postirradiation examination (PIE) measurements carried out in the hot cells determined the creep rate of zirconium alloys, and indicated that the grid plate material has hardened but has still enough residual ductility. This paper presents salient features of the design and implementation of these irradiation experiments in FBTR and the results obtained during PIE.