Zircaloy, used both as a structural material and as a fuel and poison element cladding in water-cooled reactor cores, is subject to a strain-controlled deformation, with strains in many instances, going into the plastic range. Equipment has been designed and constructed for in-reactor tensile tests in a 282 °C (540 °F) water loop at strain rates ranging from 5 × 10-6 to 10-3 h. Tensile tests of material in the unirradiated and post-irradiated conditions were made at 282 °C and 315 °C and strain rates from 3 × 10° to 5 × 10-6 h. Specimens were of longitudinal and transverse orientation to the rolling direction of a Zircaloy-4 plate. A total of seven in-reactor tests at 282 °C were conducted; two longitudinal specimens at strain rates of 10-5 and 5 × 10-6 h and five transverse specimens at strain rates from 1 × 10-4 to 5 × 10-6 h. Most specimens were tested to between 1 and 3 percent total strain. The results of the in-reactor tests to date indicate that in a neutron environment in the range of strain rates investigated, Zircaloy-4 has an apparent strain rate sensitivity index, m = 0,23, compared to unirradiated and post-irradiated Zircaloy-4 with m ≈ 0.03. As with unirradiated Zircaloy-4, the in-reactor 0.2% yield strength in the transverse direction was higher than in the longitudinal direction. The in-reactor 0.2% yield strength at 5 × 10-6 h was significantly lower than for a comparable post-irradiated test. © 1969.
