Irradiation behaviors of two novel single-phase bcc-structure high-entropy alloys for accident-tolerant fuel cladding

High-entropy alloys(HEAs)are potential alternative materials for accident-tolerant fuel cladding due to their excellent irradiation resistance and high-temperature corrosion resistance.In this work,two novel body-centered cubic(bcc)structured Mo_(0.5)Nb Ti VCr_(0.25)and Mo_(0.5)Nb Ti V_(0.5)Zr_(0.25)HEAs were fabricated.Helium-ion irradiation was performed on the two HEAs to simulate neutron irradiation,and the crystal structure,hardness,and microstructure evolution were investigated.The crystal structure of the Mo_(0.5)NbTiVCr_(0.25)HEA remained stable at low fluences,while amorphization may occur at high fluences in the two HEAs.The irradiation hardening value of the Mo_(0.5)NbTiVCr_(0.25)was 0.77 GPa at fluences of 1×10^(17)ions/cm^(2)and 1.49 GPa at fluences of 5×10^(17)ions/cm^(2),while the hardening value of the Mo_(0.5)NbTiV_(0.5)Zr_(0.25)was 1.36 GPa at ion fluences of 5×10^(17)ions/cm^(2).In comparison with most of the conventional alloys,the two HEAs showed slight irradiation hardening.The helium bubbles and dislocation loops with small size were observed in the two HEAs after irradiation.This is the first time to report the formation of a dislocation loop in bcc-structure HEAs after irradiation.However,voids and precipitates were not observed in the two HEAs which could be ascribed to the high lattice distortion and compositional complexity of HEAs.This research revealed that the two HEAs show outstanding irradiation resistance,which may be promising accident-tolerant fuel cladding materials.

The effect of Laves phase on heavy-ion radiation response of Nb-containing FeCrAl alloy for accident-tolerant fuel cladding

As a promising candidate material for the accident tolerant fuel cladding in light water reactors,the Nb-containing FeCrAl alloy has shown outstanding out-of-pile service performance due to the Laves phase precipitation.In this work,the radiation response in FeCrAl alloys with gradient Nb content under heavy ion radiation has been investigated.The focus is on the effect of the Laves phase on irradiation-induced defects and hardening.We found that the phase boundary between the matrix and Laves phase can play a critical role in capturing radiation defects,as verified by in-situ heavy-ion radiation experiments and molecular dynamic simulations.Additionally,the evolution of Laves phase under radiation is analyzed.Radiation-induced amorphization and segregations observed at high radiation doses will deepen the fundamental understanding of the stability of Laves phases in the radiation environment.

First-principles study on the diffusion behavior of Cs and I in Cr coating

Cs and I can migrate through fuel-cladding interfaces and accelerate the cladding corrosion process induced by the fuel-cladding chemical interaction.Cr coating has emerged as an important candidate for mitigating this chemical interaction.In this study,first-principles calculations were employed to investigate the diffusion behavior of Cs and I in the Cr bulk and grain boundaries to reveal the microscopic interaction mitigation mechanisms at the fuel-cladding interface.The interaction between these two fission products and the Cr coating were studied systematically,and the Cs and I temperature-dependent diffusion coefficients in Cr were obtained using Bocquet’s oversized solute-atom model and Le Claire’s nine-frequency model,respectively.The results showed that the Cs and I migration barriers were significantly lower than that of Cr,and the Cs and I diffusion coefficients were more than three orders of magnitude larger than the Cr self-diffusion coefficient within the temperature range of Generation-IV fast reactors(below 1000 K),demonstrating the strong penetration ability of Cs and I.Furthermore,Cs and I are more likely to diffuse along the grain boundary because of the generally low migration barriers,indicating that the grain boundary serves as a fast diffusion channel for Cs and I.

Segregation behavior and embrittling effect of lanthanide La, Ce, Pr,and Nd at Σ3(111) tilt symmetric grain boundary in α-Fe

The migration of lanthanide fission products to cladding materials is recognized as one of the key causes of fuel–cladding chemical interaction(FCCI) in metallic fuels during operation. We have performed first-principles density functional theory calculations to investigate the segregation behavior of lanthanide fission products(La, Ce, Pr, and Nd) and their effects on the intergranular embrittlement at Σ3(111) tilt symmetric grain boundary(GB) in α-Fe. It is found that La and Ce atoms tend to reside at the first layer near the GB with segregation energies of-2.55 eV and-1.60 eV, respectively,while Pr and Nd atoms prefer to the core mirror plane of the GB with respective segregation energies of-1.41 eV and-1.50 eV. Our calculations also show that La, Ce, Pr, and Nd atoms all act as strong embrittlers with positive strengthening energies of 2.05 eV, 1.52 eV, 1.50 eV, and 1.64 eV, respectively, when located at their most stable sites. The embrittlement capability of four lanthanide elements can be determined by the atomic size and their magnetism characters. The present calculations are helpful for understanding the behavior of fission products La, Ce, Pr, and Nd in α-Fe.

Effect of niobium content on irradiation microstructure and hardening in FeCrAl-based alloys

Iron-chromium-aluminum(FeCrAl)alloys with different content of niobium(Nb)—0,0.4 wt%,0.8 wt%,and 1.2 wt%—were designed and prepared.All samples were then irradiated with 2.4 MeV Fe^(2+)ion to the dose of 1 and 15 displacements per atom(dpa)at 400℃.The formations of dislocation loops induced by self-ion irradiation in these alloys were investigated by transmission electron microscopy(TEM).Nano-indentation tests were used to assess the hardness and irradiation hardening of samples.For the samples before irradiation,the(Fe,Cr)_(2)(Nb,Mo)Laves phases density and the nano-indentation hardness increased with increasing Nb content of the samples.After irradiation to 1 and 15 dpa,both of a/2<111>and a<100>dislocation loops were produced but no voids orα’phase were found in all samples.With increasing Nb content of the samples,the size of dislocation loops increased first and then decreased,while the total volume number density decreased and then increased.The fraction of a<100>dislocation loops increased first and then decreased with increasing Nb content,and increased with increasing irradiation dose.Dislocation networks and the amorphization of the Laves phases were observed in the samples with irradiation dose of 15 dpa.Irradiation hardening of Nb free samples was two to four times that of Nb containing samples,and the irradiation hardening increased with increasing Nb content of Nb containing samples.The experimental results indicate that the increase of Nb content in Fe Cr Al alloys can increase the density of Laves phases,leading to the decrease of Mo content and increase of Cr content in the matrix.The competition between the two types of solutes affects the nucleation and growth of the dislocation loops.

Microstructural evolution of a silicon carbide-carbon coated nanostructured ferritic alloy composite during in-situ Kr ion irradiation at 300℃ 450℃

This work focuses on irradiation behaviors of a novel silicon carbide and carbon coated nanostructured ferritic alloy(SiC-C@NFA)composite for potential applications as a cladding and structural material for next generation nuclear reactors.The SiC-C@NFA samples were irradiated with 1 MeV Kr ions up to 10 dpa at 300 and 4500 C.Microstructures and defect evolution were studied in-situ at the IVEM-Tandem facility at Argonne National Laboratory.The effects of ion irradiation on various phases such asα-ferrite matrix,(Fe,Cr)_(7)C_(3),and(Ti,W)C precipitates were evaluated.Theα-ferrite matrix showed a continuous increase in dislocation density along with spatial ordering of dislocation loops(or loop strings)at>5 dpa.The size of the dislocation loops at 450℃was larger than that at 300℃.The nucleation and growth of new(Ti,W)C precipitates inα-ferrite grains were enhanced with the ion dose at 450℃.This study provides new insight into the irradiation resistance of the SiC-C@NFA system.