Topic
Zirconium alloy
About: Zirconium alloy is a research topic. Over the lifetime, 6548 publications have been published within this topic receiving 78954 citations. The topic is also known as: zircaloy.
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Abstract: During neutron irradiation, both interstitial and vacancy loops are formed in high concentration in zirconium alloys Due to this high density of loops, the material is considerably hardened, but the recovery of the radiation damage during a heat treatment leads to a progressive softening of the irradiated material The recovery of the radiation induced hardening has been investigated using microhardness tests Transmission electron microscopy (TEM) observations performed on irradiated foils have also shown that the loop density falls while the loop size increases during the thermal annealing Furthermore, the TEM analysis has revealed that only vacancy loops are present in the material after long term annealing, the interstitial loops having entirely disappeared A numerical cluster dynamic modeling has also been used in order to reproduce the material recovery for various annealing conditions The microstructural evolution during mechanical testing with various loading conditions has also been studied It has been shown that during a creep test with low applied stress (130 MPa) and high temperature (450°C), the microstructure evolution can essentially be explained by the thermal recovery of the loops leading to glide of dislocations as found for an non-irradiated material At intermediate temperature (400°C), it is shown that for low stress level (130 MPa) the microstructure evolution can also be explained by the thermal recovery of loops, whereas for higher stress (250 MPa), sweeping of loops by gliding dislocations can also occur In addition, for an applied stress of 130 MPa and a temperature of 400°C, dislocation density is higher in the irradiated material than in the non-irradiated material deformed in the same conditions It is also shown that secondary slip systems are more activated in the irradiated material than in the non-irradiated material From this detailed analysis, the mechanical behavior during creep is interpreted in terms of microscopic deformation mechanisms
38 citations
TL;DR: In this article, the circumferential mechanical properties of zirconium alloy cladding with a variation of the hydrogen content were investigated with a ring tension test and burst test for the hydrogen charged specimens.
38 citations
TL;DR: In this article, high-energy synchrotron X-ray diffraction was used to investigate the isothermal precipitation of δ-hydride platelets in Zircaloy-4 at a range of temperatures relevant to reactor conditions, during both normal operation and thermal transients.
38 citations
TL;DR: In this article, the authors showed that the thickness of a prior-oxide layer formed on Zircaloy-4 fuel cladding can decrease during the first seconds at very high-temperature, before regrowing.
38 citations
TL;DR: In this paper, the formation of the homogeneous double oxyhydroxide of tungsten and zirconium ions acts synergistically in improving the corrosion resistance of W-Zr alloys in HCl solutions open to air at 30°C.
38 citations