Effect Of Hydrogen Absorbing And Irradiation On The Mechanical Properties Of Zircaloy-2 Cladding

Zr-2 alloy was used as cladding materials extensively in a large number of nuclear reactors. After a long-term operation at high temperature, high pressure and neutron irradiation, the material was degraded, e. g., corrosion, distortion, swelling, irradiation growth and embrittlement were observed, especially the delayed hydrogen embrittlement due to hydrogen picking up and long-term applied stress, which was one of the main factors affecting reactor safety. In this paper, the true stress-true strain relation and fracture morphology analysis were studied besides tensile testing, to evaluate the fuel cladding behavior in reactors scientifically. The data of structure and the performance can be fed back to fuel assembly design, manufacture and operating department for evaluating the life of reactors, improving the quality of fuel cladding, prolonging the burnup and improving safety abundance.Artificial hydrogen charging was adopted to achieve different hydrogen contents in Zr-2 alloy cladding. Tensile test was carried out for different hydrogen samples as well as specimens irradiated for 83 full power days in reactor. The stress-strain curve and the true stress-true strain relation were obtained. Fracture morphologies were also analyzed by Scanning Electron Microscopy(SEM).The results were evaluated as following:When the hydrogen content was less than 150μg/g, the mechanical performance of Zr-2 cladding did not change and the tensile fracture shown plastic character. At the range of 250～500μg/g, the yield stress and tensile stress increased a little accompanied by marked decreasing of elongation, which typically shown plasticity reduction caused by hydrogen picking up. When the hydrogen content reached 600μg/g, both strength and elongation decreased obviously. The total elongation was dropped to less than 1.4%, decreasing by 95% compared with original specimens, which shown obviously hydrogen embrittlement behaviors. Metallographic observation results shown that worst hydride orientation was radial.The tensile stressσ_b and yield stressσ_s were increased by 38.0% and 89.1%, respectively, and the total elongation and uniform elongation were decreased by 83.8% and 89.8% at room temperature, compared with unirradiated specimens. At 350℃, the increase ofσ_b andσ_s were 72.4% and 131.6%, and the decrease of elongation were 87.4% and 97.3%, respectively, which shown great irradiation induced hardening behaviors with the increase of strength and decrease of plasticity. It is also can be seen thatσ_s/σ_b reached almost 1, that is to say theσ_s reached saturation. The results also show that temperature have great effect onσ_s andσ_b, and a little effect on elongation.At room temperature and 350℃, the strength of samples cut from high neutron flux section was higher than that from low neutron flux section, the elongation was reversed. All fracture shown dimple character which became flat with the increase of neutron flux.The study on true stress-true strain shown that Hooke's lawσ_t=K₁ε_e for elastic stage was applicable, and Hollomon experience formulaσ_t=K₂ε_pⁿ can be used in uniform plastic stage for either unirradiated, hydrogen or irradiated specimens. Exponent n reflected the capability of uniform deforming depending on induration before necking. Exponent n decreased by 1 order due to induration.At 350℃, the Young's modulus of elasticity (E) of Zr-2 cladding is 54% of that at room temperature. The E of hydrogen samples is similar to that of original samples. Compared with the unirradiated samples, the E of irradiated specimens increased by 26% and 61%, repectively.