The ductility of highly irradiated Zircaloy-4 material was evaluated by conducting tube burst, tube tensile, and ring tensile tests on fuel cladding and guide tubes irradiated in two PWRs. The specimen fluence ranged between 9 and 12.3 × 1021 n/cm2 (E > 1 MeV), and test temperatures ranged from 313 to 673 K. The average thickness of the waterside oxide layer on the specimens ranged from 12 to 114 μm. Specimens with an oxide thickness greater than about 100μm contained regions of spalling oxide and local areas of oxide significantly thicker than the specimen average. The corresponding average hydrogen contents ranged from 40 to 674 ppm for specimens without spalling oxide and estimated to be greater than 950 ppm with spalling. Non-uniform hydride distributions were observed in the specimens due to temperature gradients during operation. The residual ductility for these high-fluence specimens is on the order of 1% uniform plastic strain for all the specimens except for two specimens with average concentration of hydrogen greater than 700 ppm and spalled oxide. The reduction in the material ductility due to radiation damage appears to be synergistically affected by localized hydride distributions and the orientation of hydride precipitates relative to the loading direction. The extent of ductility reduction due to hydride precipitates appears to be the most for the burst tests among the three tests investigated. The tensile specimens showed different fracture modes depending on deformation temperature, hydrogen concentration, local hydride volume fraction, and hydride orientation. Neckdown and spiral fractures were observed. Examination of fracture surfaces indicated ductile failure in the metallic ligaments separating the hydride precipitates that appeared to have failed in a brittle fashion. The ductility data are analyzed by treating the material as a composite of relatively ductile metal phase separated by more brittle hydride platelets. Localization of hydride phase with a reduced presence of metallic ligaments in the composite results in reduction of ductility. A local hydride volume fraction greater than a critical value is needed to initiate and propagate fracture across the specimen cross section to thereby reduce the ductility below a set value. A model is proposed to suggest a possible ductility minima at intermediate fluences and the effect of hydride precipitates on ductility.

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To establish the mechanical response of Zircaloy cladding under thermal shock conditions typical of hypothetical loss-of-coolant accident (LOCA) situations in light-water reactors (LWRs), cladding specimens were ruptured in steam during transient heating (10 K/s), oxidized at maximum temperatures between 1140 and 1770 K for various times, and cooled from the isothermal oxidation temperature to ∼1100 K at a rate of 5 K/s, and rapidly quenched by bottom flooding with water at a rate of ∼0.05 m/s. Failure “maps” for fracture of the cladding by thermal shock were developed relative to the maximum oxidation temperature and various time-dependent oxidation parameters. In situ pendulum-load impact tests were conducted at room temperature on tubes that survived the thermal quench. Information on the total absorbed energy from these tests was correlated with more extensive results from instrumented drop-weight impact tests. The thermal shock results indicate that the present Zircaloy embrittlement criterion (that is, a total oxidation limit of 17 percent of the wall thickness and a maximum cladding temperature of 1477 K) is conservative and that a more quantitative criterion, based upon the mechanical behavior of the oxidized material, can be formulated with a specified degree of conservatism consistent with the mechanical loads imposed on the cladding during reflood and the maximum amount of oxidation set by the margin of performance of emergency core-cooling systems in LWRs.

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