SEDL / STP / STP939-EB / STP28110S

Effect of Irradiation at 588 K on Mechanical Properties and Deformation Behavior of Zirconium Alloy Strip

Morize, P
Engineer, CEA Saclay, IRDI, D. Tech. SRMA, GIF sur Yvette Cedex,

Baicry, J
Engineer, CEA Saclay, IRDI, D. Tech. SRMA, GIF sur Yvette Cedex,

Mardon, JP
Material Engineer, FRAGEMA, Lyon Cedex 3,

Pages: 19    Published: Jan 1987

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Abstract

Irradiation tests conducted for the Advanced Fuel Assembly (AFA) grid development allowed the measurement of the mechanical characteristics, growth, and relaxation of the following materials: stress-relieved, α-annealed, and β-quenched Zircaloy-4, and α-annealed Zr-3Sn-1Mo. The mechanical characteristics of the zirconium alloys approached a limit value as early as 5 × 10²⁴ n m⁻², except for the β-quenched Zircaloy-4 for which the limit value was reached later at about 5 × 10²⁵ n m⁻². Uniform elongation at 588 K showed a minimum at about 5 × 10²⁴ n m⁻²; beyond this fluence, stress-relieved Zircaloy-4 presented the most marked tendency to increase. Longitudinal direction growth of annealed Zircaloy-4 and longitudinal and transverse direction growth of stress-relieved Zircaloy-4 may be expressed by the formula ε = A(φt)ⁿ where n = 0.67 (stress-relieved Zircaloy-4) and n = 0.4 (annealed Zircaloy-4). It is doubtful that the growth is associated with a density change. The stress-relaxation of the zirconium base materials is nearly complete for 4 × 10²⁵ n m⁻², with the exception of Zr-1Nb which has an intermediate behavior between those of Zircaloy-4 and Inconel 718.

Tests on elementary grid cells, beyond 2 × 10²⁵ n m⁻², showed that the load exerted by the springs of Zircaloy-4 prototype cells is completely relaxed; for the AFA cells the residual load after irradiation at about 5 × 10²⁵ n m⁻² is comparable to that observed for an Inconel 718 manufactured cell.