Influence of element substitutions on poisoning behavior of ZrV_(2)alloy:theoretical and experimental investigations

A ZrV_(2)alloy is typically susceptible to poisoning by impurity gases,which causes a considerable reduction in the hydrogen storage properties of the alloy.In this study,the adsorption characteristics of oxygen on ZrV_(2)surfaces doped with Hf,Ti,and Pd are investigated,and the effect of oxygen on the hydrogen storage performance of the alloy was discussed.Subsequently,the adsorption energy,bond-length change,density of states,and differential charge density of the alloy before and after doping are analyzed using the first-principles method.The theoretical results show that Ti doping has a limited effect on the adsorption of oxygen atoms on the ZrV_(2)surface,whereas Hf doping decreases the adsorption energy of oxygen on the ZrV_(2)surface.Oxygen atoms are more difficult to adsorb at most adsorption sites on Pd-substituting surfaces,which indicates that Pd has the best anti-poisoning properties,followed by Hf.The analysis of the differential charge density and partial density of states show that the electron interaction between the oxygen atom and surface atom of the alloys is weakened,and the total energy is reduced after Hf and Pd doping.Based on theoretical calculations,the hydrogen absorption kinetics of ZrV_(2),Zr_(0.9)Hf_(0.1)V_(2),and Zr(V_(0.9)Pd_(0.1))_(2) alloys are studied in a hydrogen-oxygen mixture of 0.5 vol%O_(2) at 25℃.The experimental results show that the hydrogen storage capacities of ZrV_(2),Zr_(0.9)Hf_(0.1)V_(2),and Zr(V_(0.9)Pd_(0.1))_(2) decrease to 19%,69%,and 80%of their original values,respectively.The order of alloy resistance to 0.5 vol%O_(2) poisoning is Zr(V_(0.9)Pd_(0.1))_(2)>Zr_(0.9)Hf_(0.1)V_(2)>ZrV_(2).Pd retains its original hydrogen absorption performance to a greater extent than undoped surfaces,and it has the strongest resistance to poisoning,which is consistent with previous theoretical calculations.

Recent studies on potential accident-tolerant fuel-cladding systems in light water reactors

Accident-tolerant fuel(ATF)has attracted considerable research attention since the 2011 Fukushima nuclear disaster.To improve the accident tolerance of the fuel-cladding systems in the current light-water reactors,it is proposed to develop and deploy(1)an enhanced Zrbased alloy or coated zircaloy for the fuel cladding,(2)alternative cladding materials with better accident tolerance,and(3)alternative fuels with enhanced accident tolerance and/or a higher U density.This review presents the features of the current UO2-zircaloy system.Different techniques and characters to develop coating materials and enhanced Zr-based alloys are summarized.The features of several selected alternative fuels and cladding materials are reviewed and discussed.The neutronic evaluations of alternative fuel-cladding systems are analyzed.It is expected that one or more types of ATF-cladding systems discussed in the present review will be implemented in commercial reactors.

Interdiffusion behavior between Cr and Zr and its effect on the microcracking behavior in the Cr-coated Zr-4 alloy

High-temperature chromium(Cr)-zirconium(Zr)interdiffusion commonly occurs in Cr-coated zircaloys applied for enhanced accident-tolerant fuel(ATF)claddings.Such interdiffusion changes the interfacial microstructure and thus the fracture mechanism of the coating under external loading.In this study,the interdiffusion behavior in a magnetron sputtered Cr coating deposited on a Zr-4 alloy was studied in a vacuum environment at 1160C.In addition,the effect of interdiffusion on the microcracking behavior of the Cr coating was determined by in situ three-point bending tests.The experimental results show that the interdiffusion behavior resulted in the formation of a ZrCr2 layer,accompanied by the consumption of Cr coating and interfacial roughening.The growth of the diffusion layer followed a nearly parabolic law with respect to annealing time,and the residual stress of the annealed coating decreased with increasing annealing time.Under external loading,a large number of cracks were generated in the brittle interlayer,and some interfacial cracks were formed and grew at the ZrCr2/Zr-4 interface.Despite the remarkable microcracks in the ZrCr2 layer,the vacuum-annealed Cr coating has significantly fewer cracks than the original coating,mainly because of the recrystallization of the coating during annealing.

Behaviors of fine(IG-110)and ultra-fine(HPG-510)grain graphite irradiated by 7 MeV Xe^26+ions

Developing a molten salt reactor needs molten salt-impermeable nuclear graphite. Ultra-fine grain graphite is a good choice as it is better in permeability than fine grain graphite. In this paper, ultra-fine grain graphite(HPG-510) and fine grain graphite(IG-110) samples are irradiated at room temperature by 7 MeV Xe ions to doses of 1 × 10^(14)-5 × 10^(15) ions/cm^2. Scanning electron microscopy, transmission electron microscopy(TEM), Raman spectroscopy and nano-indentation are used to study the radiation-induced changes. After irradiation of different doses, all the HPG-510 samples show less surface fragment than the IG-110 samples. The TEM and Raman spectra,and the hardness and modulus characterized by nano-indentation, also indicate that HPG-510 is more resistant to irradiation.