Institute of Metals Division - The Effect of Neutron Irradiation on the Tensile Properties of the Zr-2.5 Wt Pct Nb (Cb)-0.5 Wt Pct Cu Alloy

The effect of neutron irradiation on tensile properties of the Zr-2.5 wt pet Nb-0.5 wt pet Cu alloy has been evaluated for integrated neutron fluxes up to 3 x 1020 n per sq cm (E > 1 Mev). Specimen temperature during irradiation was normally 300°C. The material was studied both in annealed and in quenched and aged conditions. When the alloy (fab-ricatedfrom sponge zirconium) is quenched from a temperature =40°C below the a + ß/ß transformation and aged 6 hr at 535oC, neutron irradiation up to the maximum studied has little effect 071 reduction in area, although there is a drop in uniform elongation and the yield strength is increased by = 25 pct. Conversely, irradiation of the alloy in an annealed condition can result in much greater changes in properties, with yield-strength increases of up to 200 pet. The dose dependence of irradiation hardening obeys the saturation equntion proposed by Makin and Minter to explain hardening in copper and nickel: Act = C[1 - exp(-Døt)/1/2 A rnodel is proposed to explain the effect of niobium content and metallurgical condition on irradiatiotz behavior. THE Zr-Nb alloys form a class of material susceptible to marked strengthening by quench and age heat treatment. With niobium concentrations as low as 2.5 wt pet, strengths can be developed which are double those of annealed zircaloy-2.1 Since the neutron-capture cross section of the Zr-2.5 wt pet Nb alloy is nearly identical to that of Zircaloy-2, significant gains in power-reactor neutron economy could be obtained by replacing stressed in-reactor components of Zircaloy-2 with the heat-treated Zr-Nb alloy. When the solution heat-treatment temperature is in the high (a + ß) phase, then the tensile properties of the quenched (and aged at 500°C) material have been shown to be relatively insensitive to neutron irradiation.1-3 Addition of small quantities of copper to the binary Zr-2.5 wt pet Nb alloy gives a useful reduction to the corrosion rate in air and carbon di- oxide,&apos; and a ternary alloy of composition Zr-2.5 wt pet Nb-0.5 wt pet Cu was chosen for one component in the Douglas Point Reactor. This application utilizes both the corrosion resistance to a moist air-carbon dioxide environment and the strength developed by a quench and age heat treatment. By suitable choice of solution heat treatment and aging temperatures, tensile properties of the ternary alloy can be made nearly identical to those considered as optimum for the binary alloy. One effect of the copper addition, however, is to give marked changes in the aging kinetics of the alloy.&apos; For this reason, it appeared that neutron irradiation might promote overaging in the ternary alloy, particularly at the elevated service temperature (=300°C). This paper describes the effect of neutron irradiation on the tensile behavior of the Zr-2.5 wt pet Nb-0.5 wt pet Cu alloy in several metallurgical conditions pertinent to its use as a reactor material. The metallurgical conditions included those for which aging occurs in the unirradiated material, and fully annealed conditions for which a higher degree of thermal stability could be expected. 1) EXPERIMENTAL Two batches of alloy were used, both made from sponge zirconium. The fabricator&apos;s analysis of principle constituents showed no significant difference between the two batches, Table I. The rod material was rolled to 1.5 in. diam, and then swaged to final sizes of 0.5 to 0.375 in. diam at nominal temperatures of 785" and 625°C for the AM and AS batches, respectively. Subsequent hydrogen analyses at CRNL confirmed that the hydrogen concentration of the as-received rod was indeed <20 ppm. However, it was found that the (a + ß)/ß transus for the AM material was approximately 25°C lower than that of the AS material, indicating rather more difference in oxygen concentration than given in Table I. Except for a small difference in yield-point behavior, the tensile prop-