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Jiang, Weilin

Publications and source records attributed to Jiang, Weilin.

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Effects of irradiation temperature on the microstructure and deuterium retention in γ-LiAlO 2 pellets

This report presents the experimental findings obtained from November 2023 to September 2024. The study aims to investigate the effects of temperature on the microstructure, deuterium (D) retention, and lithium (Li) loss in γ-LiAlO 2 pellets subjected to sequential He + and D + ion irradiation to a high dose. Sequential ion irradiation has been a key method in our previous studies to simulate the behavior of γ-LiAlO 2 under neutron irradiation. In addition to dose and dose rate, irradiation temperature is a critical factor influencing microstructural and compositional changes. LiAl 5 O 8 precipitates have been observed in γ-LiAlO 2 pellets irradiated with reactor neutrons at ~300°C. These precipitates also form during ion irradiation at an elevated temperature of 500°C, but not at 300°C. In our ion irradiation experiments, the dose rate is typically three orders of magnitude higher than that of neutron irradiation, leading to a more rapid damage production. To better emulate the microstructural features in neutron-irradiated pellets using ion irradiation, a higher irradiation temperature is needed to accelerate the diffusion of point defects and enhance defect recovery rates, thereby compensating for the effects of the higher dose rate. The microstructural changes observed are the result of competing processes occurring during ion irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Molecular dynamics simulations of displacement cascades in LiAlO2 and LiAl5O8 ceramics

Abstract Molecular dynamics was employed to investigate the radiation damage due to collision cascades in LiAlO 2 and LiAl 5 O 8 , the latter being a secondary phase formed in the former during irradiation. Atomic displacement cascades were simulated by initiating primary knock-on atoms (PKA) with energy values = 5, 10 and 15 keV and the damage was quantified by the number of Frenkel pairs formed for each species: Li, Al and O. The primary challenges of modeling an ionic system with and without a core–shell model for oxygen atoms were addressed and new findings on the radiation resistance of these ceramics are presented. The working of a variable timestep function and the kinetics in the background of the simulations have been elaborated to highlight the novelty of the simulation approach. More importantly, the key results indicated that LiAlO 2 experiences much more radiation damage than LiAl 5 O 8 , where the number of Li Frenkel pairs in LiAlO 2 was 3–5 times higher than in LiAl 5 O 8 while the number of Frenkel pairs for Al and O in LiAlO 2 are ~ 2 times higher than in LiAl 5 O 8 . The primary reason is high displacement threshold energies (E d ) in LiAl 5 O 8 for Li cations. The greater E d for Li imparts higher resistance to damage during the collision cascade and thus inhibits amorphization in LiAl 5 O 8 . The presented results suggest that LiAl 5 O 8 is likely to maintain structural integrity better than LiAlO 2 in the irradiation conditions studied in this work.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Microstructural and compositional evolutions in $γ$-LiAlO 2 pellets during ion irradiation at an elevated temperature

This study reports on the microstructural, phasic and compositional evolutions of γ-LiAlO 2 during ion irradiation. Polycrystalline γ-LiAlO 2 pellets were irradiated sequentially with He + and D 2 + ions to the same combined fluences of up to 3 × 10 17 (He + +D + )/cm 2 at 773 K. The irradiated pellets were characterized using scanning transmission electron microscopy and atom probe tomography. Surface amorphization likely due to radiolysis and planar defects as a possible precursor for formation of precipitates are created at 5 × 10 16 (He + +D + )/cm 2 , followed by the formation of nano-sized precipitates and fractures at higher doses. Spinel-like precipitates of non-stoichiometric LiAl 5 O 8 and gas-filled cavities are observed to grow with increasing dose. Faceted precipitates and rounded cavities appear at 2 × 10 17 (He + +D + )/cm 2 . With further increasing ion fluence to 3 × 10 17 (He + +D + )/cm 2 , amorphization of the precipitates takes place and micron-sized fractures appear. Surface exfoliation could occur at an extremely high ion fluence. There are compositional changes in the γ-LiAlO 2 pellets during the microstructural evolution. In the precipitate and amorphized regions, Li concentrations decrease to ~7 and 3.7 at.% from 25 at.% in γ-LiAlO 2 , respectively. This study reveals a full-cycle microstructural evolution with corresponding compositional changes in γ-LiAlO 2 pellets during ion irradiation at 773 K. In conclusion, the data could help model, assess, and predict the material performance during neutron irradiation in reactors.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Atomistic simulation of brittle-to-ductile transition in silicon carbide embedded with nano-sized helium bubbles

In this study, the tensile response of cubic silicon carbide (SiC) bulk containing cavities (voids and He bubbles) has been investigated using molecular dynamic simulations. The formation of cavities in SiC leads to a significant degradation in the mechanical properties of SiC with more influence on material fracture than initial elastic deformation. The brittle-to-ductile transition occurs in cavity-embedded SiC as the pressure in He bubbles increases. This is associated with the deformation mechanism that bond breaking at a low He bubble pressure transfers to extensive dislocation activities at a higher He bubble pressure. The cavities can effectively concentrate stress around them in the direction perpendicular to the tension, which leads to preferred cracking in the region with a higher tensile stress. The failure mechanism as revealed by this study improves understanding of property degradation in SiC that may be useful for applications of SiC in advanced nuclear energy systems.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Characterization of boundary precipitation in a heavy ion irradiated tungsten heavy alloy under the simulated fusion environment

In the concerted effort to identify materials capable of surviving the adverse environment of a fusion reactor interior, tungsten heavy alloys have been put forth as leading candidates. Experimental trials and behavioral studies have yielded positive results for their adoption by taking advantage of the alloy’s unique balance of inherently high fracture toughness and low sputter yields; yet due to their relative novelty in the fusion community, there remains a lack of understanding on the response of these materials to the extended high temperature irradiation environment of the reactor interior. Further, to alleviate this issue and provide the necessary data on the behavior of tungsten heavy alloys to the simulated fusion environment, a 90W-7Ni-3Fe alloy has been subjected to elevated temperature sequential Ni + and He + ion irradiations to mimic the expected displacement damage and gas production expected after five years of service as a plasma facing material component. Atomic-scale structural analyses and nanoscale chemical mapping have identified the formation of two distinct precipitation structures, a surface localized η-carbide and an a-W 2 C type tungsten carbide, both of which appear to originate at the bi-phase interface between W and the ductile phase. This irradiation assisted and induced precipitate formation respectively is anticipated to adversely affect these materials by selective embrittlement at bi-phase interfaces leading to a reduction in the material’s overall fracture toughness during prolonged high temperature irradiation. It is asserted that any potential exposure to C during fusion reactor operational service should be minimized as to prevent the formation of these phases.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Behavior and properties of helium cavities in ductile-phase-toughened tungsten irradiated with He + ions at an elevated temperature

Ductile-phase toughened tungsten (DPT W) composites have demonstrated a great potential for plasma-facing components in fusion power systems. These materials can retain the outstanding thermomechanical properties of W while significantly enhancing the fracture toughness. The DPT W composite in this study consists of W particles (88 wt.%) embedded in a ductile NiFeW (12 wt.%) solution matrix. Irradiation of the composite was performed with 90 keV He + ions to 1.0 × 10 17 He + /cm 2 at 973 K. The He cavities formed in both W and NiFeW phases are examined using convergent-beam scanning transmission electron microscopy (CB-STEM) at an atomic-level resolution. Full-range depth profiles of the He cavity diameters and number densities are analyzed. The results show that the average cavity diameter in NiFeW is ~2.7 times that in W, while the number density is an order of magnitude lower. A quantitative analysis of the He density and pressure in the cavities is also achieved using STEM electron energy loss spectroscopy (STEM-EELS) mapping. The He atoms in the cavities in NiFeW are found to be under-pressurized, while those in similar-sized cavities in W are not detectable. Further, the He energy-loss shift exhibits a linear relationship with the He density in cavities in NiFeW, with the slope comparable to that for martensitic steel.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Ion irradiation study of lithium silicates for fusion blanket applications

Here, this study reports on the microstructural and compositional evolutions, deuterium release and lithium loss behavior in lithium silicates. Pellets of biphasic Li 4 SiO 4 and Li 2 SiO 3 were fabricated using hot pressing of powders. Sequential irradiation with Si + , He + and D + ions was performed up to 773 K to emulate one-year 6 Li burnup in 6 Li 4 SiO 4 inside the SlimCS DEMO. Crystalline Li 4 SiO 4 phase was not observed in the irradiated depth region of all the samples in this study. In spite of full amorphization in the near-surface region, Li 2 SiO 3 irradiated to a high dose at 773 K remained crystalline in the damage peak region, suggesting that Li 2 SiO 3 is more irradiation resistant to amorphization than Li 4 SiO 4 . Radiolysis may be primarily responsible for silicate decomposition and amorphization. Data from this study also show that D release from the pellet is very efficient during ion irradiation at 773 K, which is accompanied with a significant Li loss. Thermal annealing for 10 min at 773 K for the room-temperature irradiated pellet leads to a nearly complete D release without an observable Li loss. A concept of Ni coating on tritium breeder materials is discussed with supporting data to minimize Li loss without a significant impact on tritium diffusion and release.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Fusion Blanket and Fuel Cycle Research at PNNL: FY22 Year-end Report

During the reporting period, research at PNNL focused on two tasks within the DOE Fusion Blanket and Fuel Cycle Program. Research on Task 1, Tritium Extraction from Pb-Li and He Using a Vacuum Permeator, focused on atomistic modeling to better understand tritium transport in Pd-coated V vacuum permeators. As a lower cost alternative to Pd permeators, thin coatings of Pd (or other noble metals) can be deposited over a substrate like V. However, the permeation performance of composite metal membranes degrades over time, due to the formation of intermetallics at the coating-substrate interface. Computational studies were performed to better understand tritium transport through these Pd-V intermetallics. The results of the FY22 Pd-V modeling study were recently submitted for publication in Computational Materials Science and presented at the Technology of Fusion Energy conference. Future work in this area will focus on interdiffusion barriers to prevent intermetallic formation that is deleterious to tritium transport. There are opportunities for collaboration with researchers at the Colorado School of Mines, who are manufacturing and testing candidate interdiffusion barriers. Research on Task 3, Solid Breeder Materials, included ion irradiation and post-irradiation characterization of lithium orthosilicate (Li 4 SiO 4 ) and lithium metasilicate (Li 2 SiO 3 ) to improve fundamental understanding of irradiation effects, in combination with atomistic modeling focused on the energetics of He clustering in these two ceramic phases. The results of the study suggested that the Li 4 SiO 4 phase, which is more desirable as a solid breeder due to its higher Li density, was amorphized during ion irradiation while the Li 2 SiO 3 phase appeared to be more resistant to irradiation damage. It is possible that Li loss contributed to the poor irradiation performance of the Li 4 SiO 4 , and some thoughts are provided regarding coatings that could be applied to solid breeders like this to prevent Li loss at elevated temperature while not hindering tritium diffusion. The results of the FY22 ion irradiation study were recently submitted for publication in Journal of Nuclear Materials and presented at the 22nd International Conference on Ion Beam Modification of Materials. Future work in this area will focus on Li-rich ceramics such as Li 5 AlO 4 and Li 8 ZrO 6 that have high Li density and should provide rapid tritium release based on previous work with less Li-rich ceramics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Microstructural evolution and precipitation in γ-LiAlO 2 during ion irradiation

Polycrystalline γ-LiAlO 2 pellets were sequentially irradiated with 120 keV He + and 80 keV D 2 + ions to the fluences of 1 × 10 17 and 2 × 10 17 (He + + D + )/cm 2 at 573 K. Additional irradiation was performed to a fluence of 2 × 10 17 (He + + D + )/cm 2 at 773 K. The irradiated pellets were characterized using scanning transmission electron microscopy, time-of-flight secondary ion mass spectrometry, and grazing incidence x-ray diffraction. Lattice damage, amorphization, and fractures are observed with no evidence for the formation of secondary-phase precipitates in the pellets irradiated up to an ion fluence of 2 × 10 17 (He + + D + )/cm 2 at 573 K. In contrast, faceted precipitates with sizes larger than 100 nm formed in a pellet irradiated to 2 × 10 17 (He + + D + )/cm 2 at 773 K. Analyses of the diffraction and composition data suggest that the precipitates have a spinel-type structure, likely a non-stoichiometric LiAl 5 O 8 with Li depletion. This could be an intermediate phase with Li atoms at the octahedral and possibly tetrahedral sites as well. It is speculated that as the dose increases, Li loss will continue and the precipitates will approach a composition of alumina primarily in phases of α-Al 2 O 3 and amorphized Al 2 O 3 .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

Behavior of helium cavities in ion-irradiated W-Ni-Fe ductile-phase toughened tungsten

This study reports on the distribution of helium (He) cavities in a hot-rolled W-Ni-Fe ductile-phase toughened tungsten (DPT W) composite irradiated to a dose and a helium concentration that are comparable to those in the material after 5-year irradiation in a conceptual fusion power plant. Here, the DPT W sample consists of W particles embedded in a ductile-phase NiFeW matrix with a nominal composition of 90W-7Ni-3Fe by weight. It was hot-rolled to a thickness reduction by 87% (87R DPT W). Sequential irradiation of the material with 1.2 MeV Ni + ions to a fluence of 2.15×10 16 Ni + /cm 2 and 90 keV He + ions to 6.5×10 15 He + /cm 2 was performed at 973 K. Larger He cavities with a lower number density are observed in NiFeW than W. Helium cavities are aggregated preferentially along the NiFeW/W interphase boundary. This behavior is not observed along the W/W grain boundary under the same irradiation conditions. A similar distribution of He cavities with comparable sizes is observed in 87R DPT W irradiated with He + ions only, suggesting that the He cavity distribution is not significantly affected by pre-irradiation of Ni + ions at 973 K. The data from this study does not provide any evidence for formation of visible voids or Ni precipitates in W phase irradiated with Ni + ions only. Diffusion and clustering of He atoms and mobile vacancies during ion irradiation at 973 K are responsible for the formation and growth of the He cavities.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Energy-Dispersive X-ray Spectroscopy and Atom-probe Tomography Data Quantifying Component-Ratios of Multicomponent Nano-Precipitates in Ion-Irradiated Ceria

Samples of ~1 µm films of CeO2 doped with 2 wt% Mo, 1.5 wt% Ru, 0.75 wt% Pd, 0.5 wt% Re and 0.25 wt% Rh grown with pulsed laser deposition were irradiated with I2+ ions (610°C and 730°C, 1016 and 5×1016 I2+/cm2). For selected samples post-irradiation heat treatment was conducted (900°C, 1100°C). The specimens were sectioned with focussed ion beam milling and characterized in a transmission electron microscope with energy dispesive x-ray spectroscopy, and with atom-probe tomography. Energy dispersive x-ray spectroscopy was used to obtain elemental maps showing the distribution of dopants in the specimen after exposure. Some of these maps are discussed in detail in our companion article “Formation of multicomponent alloy particles in doped ceria under I2+ ion irradiation and thermal annealing” in the Journal of Nuclear Materials [REF]. Advanced computational analysis could be used to more accurately quantify local compositions. Data is provided for additional regions of interest and one additional irradiation condition. The doped Ceria film that was heat treated at 1100 °C delaminated from the substrate in most places. Samples were extracted from the underside of a delaminated piece and analyzed with atom-probe tomography. The resulting data show Ceria and a Mo-rich particle and demonstrate that this approach is feasable in principle to study local compositions in a sample exposed to such extreme conditions.

multicomponent alloys, epsilon phase, EDS, atom pr↗
DOE OSTI2021

Temperature dependence of irradiation-induced nanocrystallization in amorphous silicon carbide

Ion irradiation induced nucleation and growth in amorphous SiC at elevated temperatures are investigated in this study. Both as-deposited amorphous and Kr ion irradiation-amorphized SiC films are used for the investigation. Similar behavior of particle precipitation in the two types of the films is observed. It is found that the threshold temperature for nucleation is between 550 and 700 K, which is considerably higher than the critical temperature for full amorphization in SiC irradiated with Kr ions. There is a temperature regime up to at least 900 K, where the growth rates are only weakly dependent of the irradiation temperature. This temperature regime is potentially useful to tailor the size and density of nanocrystalline SiC precipitates by optimizing the ion irradiation conditions.

Ion irradiation, Nucleation and Growth, Silicon Ca↗