Thermomechanical Behavior and Modeling Between 350°C and 400°C of Zircaloy-4 Cladding Tubes From an Unirradiated State to High Fluence (0 to 85s˙1024 nm−2,E>1 MeV)

Abstract

This paper first describes the effect of neutron irradiation on the thermomechanical behavior of stress-relieved Zircaloy-4 fuel tubes that have been analyzed after exposure to five different fluences ranging from nonirradiated material to high burnup. In the second part, a viscoplastic model is proposed to simulate, for different isotherms, 350°C<T<400°C, out-of-flux anisotropic mechanical behavior of the cladding tubes over the fluence range 0<ϕ<100s˙1024 nm−2E>1 MeV. The model, identified for tests conducted at 350°C, has been validated from tests made at 380°C and 400°C. The model is capable of simulating strain hardening under internal pressure followed by a stress relaxation period, the loading producing an interaction between the pellet and cladding. Introduction of a state variable characterizing the damage caused by a bombardment with neutrons into the model has allowed us to simulate the irradiation-induced hardening and creep rate decrease, as well as the saturation noticed after two cycles of irradiation ≅45s˙1024 nm−2E>1 MeV in a pressurized water reactor (PWR). Finally, the numerical simulations show the model is able to reproduce the totality of the thermomechanical experiments. [S0094-4289(00)00202-4]

A model to describe the anisotropic viscoplastic mechanical behavior of fresh and irradiated Zircaloy-4 fuel claddings under RIA loading conditions

“…2 and 9). The axial to hoop plastic strain ratios used in the present work are in good agreement with the approximate ratio of À1.8 reported in [3,4,15] for the non-irradiated material under axial tension at 350°C. As shown in Fig.…”

“…Several models, either based on macroscopic constitutive equations [3][4][5] or micromechanical polycrystalline descriptions [6][7][8], have been proposed in the literature to simulate the mechanical behavior of zirconium alloys cladding materials. However, most of them are not appropriate in the field of RIA studies as they are usually restricted to limited temperature and strain rate ranges specific of normal or slightly off-normal conditions.…”

A polycrystalline modeling of the mechanical behavior of neutron irradiated zirconium alloys

“…Compared to macroscopic models [43,36,76,77], this polycrystalline model has the great advantage to compute explicitly the kinematic strain hardening as a consequence of the interaction between the grains instead of a full empirical kinematic strain hardening. The isotropic strain hardening, consequence of the evolution of the various CRSS, is also modeled more physically by taking into account the activation of the different slip systems, such as the main activation of the basal slip in the case of the irradiated material.…”

Experimental Approaches for Manufacturing of Simulated Cladding and Simulated Fuel Rod for Mechanical Decladder

“…According to R. S. Daum, the mechanical properties of the investigated zircaloy-4 were reported (tensile strength 578 MPa, yield strength 706 MPa, and elongation 7%) [16]. On these bases and Table 3, we see that the mechanical properties of the actual spent fuel and the simulated cladding tube are similar [17][18][19]. On the order hand, existing mechanical property databases for zircaloy-4 include data from biaxial tube burst, axial tensile, and ring tensile tests for cladding irradiated up to about 60 GWd/MTU [20,21].…”