Irradiation hardening behaviors of tungsten-potassium alloy studied by accelerated 3-MeV W^2+ ions

Tungsten-potassium(WK)alloy with ultrafine/fine grains and nano-K bubbles is fabricated through spark plasma sintering(SPS)and rolling process.In this study,3-MeV W^2+ ion irradiation with a tandem accelerator is adopted to simulate the displacement damage caused by neutrons.As the depth of irradiation damage layer is limited to only 500 nm,the hardening behaviors of WK alloy and ITER(International Thermonuclear Experimental Reactor)-W under several damage levels are investigated through Bercovich tip nanoindentation test and other morphological characterizations.The indenter size effect(ISE),soft substrate effect(SSE),and damage gradient effect(DGE)are found to influence the measurement of nano-hardness.Few or no pop-ins in irradiated samples are observed while visible pop-in events take place in unirradiated metals.Extensive pile-up with different morphology features around the indentation exists in both WK and ITER-W.The WK shows a smaller hardness increment than ITER-W under the same condition of displacement damage.This study provides beneficial information for WK alloy serving as a promising plasma facing materials(PFMs)candidate.

Hardening effect of multi-energyW2+-ion irradiation on tungsten–potassium alloy

Tungsten is one of the most promising plasma-facing materials (PFMs) to be used in the nuclear fusion reactor as divertor material in the future. In this work, W2+-ions bombardment is used to simulate the neutron irradiation damage to commercial pure tungsten (W) and rolled tungsten–potassium (W–K). The 7 MeV of 3 × 10^15 W2+-ions/cm2, 3 MeV of 4.5 × 10^14 W2+, and 2 MeV of 3 × 10^14 W2+-ions/cm2 are applied at 923 K in sequence to produce a uniform region of 100 nm–400 nm beneath the sample surface with the maximum damage value of 11.5 dpa. Nanoindentation is used to inspect the changes in hardness and elastic modulus after self-ion irradiation. Irradiation hardening occurred in both materials. The irradiation hardening of rolled W–K is affected by two factors: one is the absorption of vacancies and interstitial atoms by potassium bubbles, and the other is the interaction between potassium bubbles and dislocations. Under the condition of 11.5 dpa, the capability of defect absorption can reach a threshold. As a result, dislocations finally dominate the hardening of rolled W–K. Specific features of dislocation loops in W–K are further observed by transmission electron microscopy (TEM) to explain the hardening effect. This work might provide valuable enlightenment for W–K alloy as a promising plasma facing material candidate.

Achieving a remarkable low-temperature tensile ductility in a high-strength tungsten alloy

Hot-swaging yields a high ultimate tensile strength of 712 MPa but a limited tensile ductility with the total elongation of3.6%at a testing temperature of 200℃in a representative W-0.5wt.%ZrC alloy.In this work,the evolution of Vickers microhardness with annealing temperatures is investigated in detail,which contributes to a rough index chart to guide the search for an optimized post-annealing temperature.Through the post-annealing around 1300℃,an outstanding tensile ductility at200℃,including a uniform elongation of 14%and a total elongation of~25%,has been achieved without the sacrifice of its strength.The evolution of dislocations and grain structures with the annealing temperatures accessed through backscattered scanning electron microscope and transmission electron microscope analysis reveals that the improved low-temperature tensile ductility has resulted from the reduction of residual dislocations and dislocation tanglement via the static recovery,which provides more room to accommodate dislocations,and hence stronger strain hardening ability and tensile ductility.

Mechanical property and thermal shock behavior of tungsten-yttria-stabilized zirconia cermet fabricated by spark plasma sintering

The cermet fuels have been considered as a potential key component for the nuclear thermal propulsion,and the homogeneity of the fuel particles in the metal matrix plays a crucial role in stabilizing the structure at extremely high temperatures.In this work,liquid paraffin was used as additive to improve the distribution of yttria-stabilized zirconia(YSZ,an appropriate surrogate for UO_(2) fuel)microspheres in the tungsten(W)matrix,and the W-YSZ cermet wafers(volume ratio 1:1)with a relative density of 97.6%were fabricated by spark plasma sintering with a specifically designed program.The effects of the paraffin dosage(0-5 wt.%)on the homogeneity,microstructure,mechanical properties,and the thermal conductivity of W-YSZ cermet were investigated.The W-YSZ sample with 2 wt.%paraffin shows the highest homogeneity and exhibits the best comprehensive properties,including the ultimate tensile strength of 132.2 MPa at 600℃,the bending strength of455 MPa and thermal conductivity of 50 W·m^(-1)·K^(-1)at room temperature.Moreover,the cermet could keep structurally sound after thermal shocked at a heat load of 20 MW·m^(-2).These results would be helpful for the design and optimization of the cermet fuels in the nuclear thermal propulsion.

A deep learning interatomic potential suitable for simulating radiation damage in bulk tungsten

So far, it has been a challenge for existing interatomic potentials to accurately describe a wide range of physical properties and maintain reasonable efficiency. In this work, we develop an interatomic potential for simulating radiation damage in body-centered cubic tungsten by employing deep potential, a neural network-based deep learning model for representing the potential energy surface. The resulting potential predicts a variety of physical properties consistent with first-principles calculations, including phonon spectrum, thermal expansion, generalized stacking fault energies, energetics of free surfaces, point defects, vacancy clusters, and prismatic dislocation loops. Specifically, we investigated the elasticity-related properties of prismatic dislocation loops, i.e., their dipole tensors, relaxation volumes, and elastic interaction energies. This potential is found to predict the maximal elastic interaction energy between two 1/2 <1 1 1> loops better than previous potentials, with a relative error of only 7.6%. The predicted threshold displacement energies are in reasonable agreement with experimental results, with an average of 128 eV. The efficiency of the present potential is also comparable to the tabulated gaussian approximation potentials and modified embedded atom method potentials, meanwhile, can be further accelerated by graphical processing units. Extensive benchmark tests indicate that this potential has a relatively good balance between accuracy, transferability, and efficiency.

First-principles study of substitutional solute and carbon interactions in tungsten

Interstitial carbon and substitutional transition metal(TM)solutes are common impurities in tungsten and tungsten alloys.Yet,despite its important role in affecting mechanical and irradiation performances of tungsten,the interplay between these impurities remains largely unknown.In this work,we performed systematic first-principles simulations to study the interaction between carbon and TM solutes.By calculating related binding energies,we found that interplay between carbon and TM solutes is dominated by elastic interactions,with carbon generally showing attractions to TM solutes.Further,including vacancies in our calculation,we found that all solute-vacancy-carbon complexes are energetically stable with respect to associated point defects.Additional analysis shows that vacancy-carbon binding is generally weakened by TM solutes,while carbon also in turn reduces the binding energy between vacancy and TM solutes.Based on these binding energy results,we,respectively,evaluated the effect of solute and carbon on each other’s diffusion behaviors.We found that Cr and V slightly decrease the carbon diffusivity while other commonly seen TM solutes show little impacts on carbon diffusion,and we also expect carbon to slow down vacancy-mediated TM solute diffusion in tungsten.

Low-temperature mechanical and magnetic properties of the reduced activation martensitic steel

Mechanical and magnetic properties as well as their relationship in the reduced activation martensitic （RAM） steel were investigated in the temperature range from -90℃ to 20℃. Charpy impact tests show that the ductile-to-brittle transition temperature （DBTT） of the RAM steel is about -60℃. Low-temperature tensile tests show that the yield strength, ultimate tensile strength and total elongation values increase as temperature decreases, indicating that the strength and plasticity below the DBTT are higher than those above the DBTT. The coercive field （Hc） in the scale of logarithm decreases linearly with the increasing temperature and the absolute value of the slope of InHc versus temperature above the DBTT is obviously larger than that below the DBTT, also confirmed in the T91 steel. The results indicate that the non-destructive magnetic measurement is a promising candidate method for the DBTT detection of ferromagnetic steels.