Yaqiao Wu

"A Preliminary Investigation of High Dose Ion Irradiation Response of a Lanthana-Bearing Nanostructured Ferritic Steel Processed via Spark Plasma Sintering" Somayeh Pasebani, Indrajit Charit, Ankan Guria, Yaqiao Wu, Jatuporn Burns, Darryl Butt, James Cole, Lin Shao, Journal of Nuclear Materials Vol. 495 2017 78-84 Link
A lanthana-containing nanostructured ferritic steel (NFS) was processed via mechanical alloying (MA) of Fe-14Cr-1Ti-0.3Mo-0.5La2O3 (wt.%) and consolidated via spark plasma sintering (SPS). In order to study the consolidation behavior via SPS, sintering temperature and dwell time were correlated with microstructure, density, microhardness and shear yield strength of the sintered specimens. A bimodal grain size distribution including both micron-sized and nano-sized grains was observed in the microstructure of specimens sintered at 850, 950 and 1050 oC for 45 min. Significant densification occurred at temperatures greater than 950 oC with a relative density higher than 98%. A variety of nanoparticles, some enriched in Fe and Cr oxides and copious nanoparticles smaller than 10 nm with faceted morphology and enriched in La and Ti oxides were observed. After SPS at 950 oC, the number density of Cr-Ti-La-O enriched nanoclusters with an average radius of 1.5 nm was estimated to be 1.2 ×10^24 m^-3 . The La + Ti : O ratio was close to 1 after SPS at 950 and 1050 C; however, the number density of nanoclusters decreased at 1050 C. With SPS above 950 C, the density improved but the microhardness and shear yield strength decreased due to partial coarsening of the grains and nanoparticles.
"Atom probe study of irradiation-enhanced a' precipitation in neutron-irradiated Fe–Cr model alloys" Wei-Ying Chen, Jian Gan, Stuart Maloy, Kun Mo, Maria Okuniewski, James Stubbins, Yinbin Miao, Yaqiao Wu, Carolyn Tomchik, Journal of Nuclear Materials Vol. 462 2015 242-249 Link
Atom probe tomography (APT) was performed to study the effects of Cr concentrations, irradiation doses and irradiation temperatures on α′ phase formation in Fe–Cr model alloys (10–16 at.%) irradiated at 300 and 450 °C to 0.01, 0.1 and 1 dpa. For 1 dpa specimens, α′ precipitates with an average radius of 1.0–1.3 nm were observed. The precipitate density varied significantly from 1.1 × 1023 to 2.7 × 1024 1/m3, depending on Cr concentrations and irradiation temperatures. The volume fraction of α′ phase in 1 dpa specimens qualitatively agreed with the phase diagram prediction. For 0.01 dpa and 0.1 dpa, frequency distribution analysis detected slight Cr segregation in high-Cr specimens, but not in Fe–10Cr specimens. Proximity histogram analysis showed that the radial Cr concentration was highest at the center of α′ precipitates. For most precipitates, the Cr contents were significantly lower than that predicted by the phase diagram. The Cr concentration at precipitate center increased with increasing precipitate size.
"Characterization of stress corrosion cracks in Ni-based weld alloys 52, 52M and 152 grown in high-temperature water" Yi Xie, Yaqiao Wu, Jatuporn Burns, Jinsuo Zhang, Materials Characterization Vol. 112 2016 87-97 Link
Ni-based weld alloys 52, 52M and 152 are extensively used in repair and mitigation of primary water stress corrosion cracking (SCC) in nuclear power plants. In the present study, a series of microstructure and microchemistry at the SCC tips of these alloys were examined with scanning electron microscopy (SEM), electron backscatter diffraction (EBSD), transmission electron microscopy (TEM), energy-dispersive X-ray spectroscopy (EDS), scanning transmission electron microscopy (STEM) and energy filtered transmission electron microscopy (EFTEM). The specimens have similar chemical compositions and testing conditions. Intergranular (IG) and transgranular (TG) SCC was observed in all of them. The cracks were filled with nickel-oxides and partial precipitations of chrome carbides (CrCs), niobium carbides (NbCs), titanium nitrides (TiNs) and silicon carbides (SiCs), while iron (Fe) was largely dissolved into the solution. However, the crack densities, lengths and distributions were different for all three specimens.
"Deformation-assisted rejuvenation of irradiation-induced phase instabilities in Cu-Ta heterophase nanocomposite" Janelle Wharry, Priyam Patki, Yaqiao Wu, B. Chad Hornbuckle, Kristopher Darling, JOM Vol. 74 2022 4094-4106 Link
"Effects of neutron irradiation and post-irradiation annealing on the microstructure of HT-UPS stainless steel" Chi Xu, Wei-Ying Chen, Xuan Zhang, Meimei Li, Yong Yang, Yaqiao Wu, Journal of Nuclear Materials Vol. 507 2018 188-197 Link
Microstructural changes resulted from neutron irradiation and post-irradiation annealing in a high-temperature ultra-fine precipitate strengthened (HT-UPS) stainless steel were characterized using transmission electron microscopy (TEM) and atom probe tomography (APT). Three HT-UPS samples were neutron-irradiated to 3 dpa at 500?°C, and after irradiation, two of them were annealed for 1?h?at 600?°C and 700?°C, respectively. Frank dislocation loops were the dominant defect structure in both the as-irradiated and 600?°C post-irradiation-annealed (PIAed) samples, and the loop sizes and densities were similar in these two samples. Unfaulted dislocation loops were observed in the 700?°C PIAed sample, and the loop density was greatly reduced in comparison with that in the as-irradiated sample. Nano-sized MX precipitates were observed under TEM in the 700?°C PIAed sample, but not in the 600?°C PIAed or the as-irradiated samples. The titanium-rich clusters were identified in all three samples using APT. The post-irradiation annealing (PIA) caused the growth of the Ti-rich clusters with a stronger effect at 700?°C than at 600?°C. The irradiation caused elemental segregations at the grain boundary and the grain interior, and the grain boundary segregation behavior is consistent with observations in other irradiated austenitic steels. APT results showed that PIA reduced the magnitude of irradiation induced segregations.
"Effects of proton irradiation on microstructure and mechanical properties of nanocrystalline Cu–10at%Ta alloy" Priyam Patki, Yaqiao Wu, Janelle Wharry, Materialia Vol. 9 2020 Link
"Enhanced Resistance to Irradiation Induced Ferritic Transformation in Nanostructured Austenitic Steels" Andrew Hoffman, Maalavan Arivu, Haiming Wen, Li He, Kumar Sridharan, Xin Wang, Xiang Liu, Lingfeng He, Yaqiao Wu, Materialia Vol. 13 2020 Link
Irradiation induced phase transformation of γ-austenite to α-ferrite has been observed in austenitic steels for the past several decades. This transformation can be detrimental to structural materials in a nuclear reactor environment as the increased fraction of the ferritic phase can increase corrosion and embrittlement and lead to stress corrosion cracking. This transformation is caused by both strain induced martensite transformation as well as radiation induced segregation and precipitation. In this study, two radiation tolerant nanostructured 304L austenitic steels (one ultrafine grained and one nanocrystalline) were manufactured using severe plastic deformation. These nanostructured 304L steels were compared to conventional coarse-grained 304L, after self-ion irradiation at 500°C up to a peak damage of 50 displacements per atom. Phase fraction after irradiation was analyzed using grazing incidence x-ray diffraction, precession electron diffraction, and electron backscatter diffraction. Nanostructured 304L steels showed significant resistance to irradiation induced austenite to ferrite transformation. This resistance was shown to be due to a decrease in defect formation, as well as a reduction in radiation induced segregation and precipitation.
"In situ TEM mechanical testing: an emerging approach for characterization of polycrystalline, irradiated alloys" Janelle Wharry, Kayla Yano, Matthew Swenson, Yaqiao Wu, Microscopy & Microanalysis Vol. 22 2016 1478 Link
"Ion irradiation and examination of Additive friction stir deposited 316 stainless steel" Priyanka Agrawal, Ching-Heng Shiau, Aishani Sharma, Zhihan Hu, Megha Dubey, Yu Lu, Lin Shao, Ramprashad Prabhakaran, Yaqiao Wu, Rajiv Mishra, Materials & Design Vol. 238 2024 112730 Link
This study explored solid-state additive friction stir deposition (AFSD) as a modular manufacturing technology, with the aim of enabling a more rapid and streamlined on-site fabrication process for large meter-scale nuclear structural components with fully dense parts. Austenitic 316 stainless steel (SS) is an excellent candidate to demonstrate AFSD, as it is a commonly-used structural material for nuclear applications. The microstructural evolution and concomitant changes in mechanical properties after 5 MeV Fe++ ion irradiation were studied comprehensively via transmission electron microscopy and nanoindentation. AFSD-processed 316 SS led to a fine-grained and ultrafine-grained microstructure that resulted in a simultaneous increase in strength, ductility, toughness, irradiation resistance, and corrosion resistance. The AFSD samples did not exhibit voids even at 100 dpa dose at 600 °C. The enhanced radiation tolerance as compared to conventional SS was reasoned to be due to the high density of grain boundaries that act as irradiation-induced defect sinks.
"Irradiation damage in (Zr0.25Ta0.25Nb0.25Ti0.25)C high-entropy carbide ceramics" Bai Cui, Fei Wang, Xueliang Yan, Lin Shao, Tianyao Wang, Yaqiao Wu, Michael Nastasi, Yongfeng Lu, Acta Materialia Vol. 2020 Link
"Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel" Todd Allen, Yiren Chen, Zhangbo Li, Wei-Yang Lo, Janne Pakarinen, Yaqiao Wu, Yong Yang, Journal of Nuclear Materials Vol. 466 2015 201-207 Link
To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ~315 °C to 0.08 dpa (5.6 × 1019 n/cm2, E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10-9 dpa/s was found to induce spinodal decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.
"Lanthana-bearing nanostructured ferritic steels via spark plasma sintering" SULTAN ALSAGABI, Darryl Butt, Indrajit Charit, James Cole, Somayeh Pasebani, Yaqiao Wu, Jatuporn Burns, Kerry Allahar, Journal of Nuclear Materials Vol. 470 2016 297-306 Link
A lanthana-containing nanostructured ferritic steel (NFS) was processed via mechanical alloying (MA) of Fe-14Cr-1Ti-0.3Mo-0.5La2O3 (wt.%) and consolidated via spark plasma sintering (SPS). In order to study the consolidation behavior via SPS, sintering temperature and dwell time were correlated with microstructure, density, microhardness and shear yield strength of the sintered specimens. A bimodal grain size distribution including both micron-sized and nano-sized grains was observed in the microstructure of specimens sintered at 850, 950 and1050 °C for 45 min. Significant densification occurred at temperatures greater than 950 °C with a relative density higher than 98%. A variety of nanoparticles, some enriched in Fe and Cr oxides and copious nanoparticles smaller than 10 nm with faceted morphology and enriched in La and Ti oxides were observed. After SPS at 950 °C, the number density of Cr–Ti–La–O-enriched nanoclusters with an average radius of 1.5 nm was estimated to be 1.2 × 1024 m−3. The La + Ti:O ratio was close to 1 after SPS at 950 and 1050 °C; however, the number density of nanoclusters decreased at 1050 °C. With SPS above 950 °C, the density improved but the microhardness and shear yield strength decreased due to partial coarsening of the grains and nanoparticles.
"Laser weld-induced formation of amorphous Mn-Si precipitate in 304 stainless steel" Janelle Wharry, Keyou Mao, Yaqiao Wu, Cheng Sun, Emmanuel Perez, Materialia Vol. 3 2018 174-177 Link
"Materials qualification through the Nuclear Science User Facilities (NSUF): A case study on irradiated PM-HIP structural alloys" Janelle Wharry, Donna Guillen, Caleb Clement, Saquib Bin Habib, Wen Jiang, Yu Lu, Yaqiao Wu, Ching-Heng Shiau, David Frazer, Brenden Heidrich, Collin Knight, David Gandy, Frontiers in Nuclear Engineering Vol. 2 2023 1306529 Link
"Mechanical Alloying of Lanthana-Bearing Nanostructured Ferritic Steels" Darryl Butt, James Cole, Somayeh Pasebani, Yaqiao Wu, Indrajit Charit, Acta Materialia Vol. 61 2013 5605-5617 Link
A novel nanostructured ferritic steel powder with the nominal composition Fe–14Cr–1Ti–0.3Mo–0.5La2O3 (wt.%) was developed via high energy ball milling. La2O3 was added to this alloy instead of the traditionally used Y2O3. The effects of varying the ball milling parameters, such as milling time, steel ball size and ball to powder ratio, on the mechanical properties and microstructural characteristics of the as-milled powder were investigated. Nanocrystallites of a body-centered cubic ferritic solid solution matrix with a mean size of approximately 20 nm were observed by transmission electron microscopy. Nanoscale characterization of the as-milled powder by local electrode atom probe tomography revealed the formation of Cr–Ti–La–O-enriched nanoclusters during mechanical alloying. The Cr:Ti:La:O ratio is considered “non-stoichiometric”. The average size (radius) of the nanoclusters was about 1 nm, with number density of 3.7x10^24 m^-3. The mechanism for formation of nanoclusters in the as-milled powder is discussed. La2O3 appears to be a promising alternative rare earth oxide for future nanostructured ferritic steels.
"Mechanical testing data from neutron irradiations of PM-HIP and conventionally manufactured nuclear structural alloys" Donna Guillen, Janelle Wharry, Caleb Clement, Yangyang Zhao, Katelyn Wachs, David Frazer, Jatuporn Burns, Yu Lu, Yaqiao Wu, Collin Knight, David Gandy, Data in Brief Vol. 48 2023 109092 Link
This article presents the comprehensive mechanical testing data archive from a neutron irradiation campaign of nuclear structural alloys fabricated by powder metallurgy with hot isostatic pressing (PM-HIP). The irradiation campaign was designed to facilitate a direct comparison of PM-HIP to conventional casting or forging. Five common nuclear structural alloys were included in the campaign: 316L stainless steel, SA508 pressure vessel steel, Grade 91 ferritic steel, and Ni-base alloys 625 and 690. Irradiations were carried out in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to target doses of 1 and 3 displacements per atom (dpa) at target temperatures of 300 and 400 °C. This article contains the data collected from post-irradiation uniaxial tensile tests following ASTM E8 specifications, fractography of these tensile bars, and nanoindentation. By making this systematic and valuable neutron irradiated mechanical behavior dataset openly available to the nuclear materials research community, researchers may now use this data to populate material performance databases, validate material performance and hardening models, design follow-on experiments, and enable future nuclear code-qualification of PM-HIP techniques.
"Method for Evaluating Irradiation Effects on Flow Stress in Fe-9%Cr ODS Using TEM In Situ Cantilevers" Kayla Yano, Yaqiao Wu, Janelle Wharry, Journal of Minerals, Metals & Materials Society Vol. 72 2020 2065-2074 Link
"Method for Fabricating Depth-Specific TEM in situ Tensile Bars" Patrick Warren, Yaqiao Wu, Janelle Wharry, George Warren, Megha Dubey, Jatu Burns, Journal of Materials Vol. 72 [unknown] 2057 - 2064 Link
The growing use of ion irradiation to assess degradation of nuclear materials has created a need to develop novel methods to probe the mechanical response of shallow ion-irradiated layers. Transmission electron microscopy (TEM) in situ mechanical testing can isolate the ion-irradiated layer from its unirradiated substrate. However, there is a lack of established procedures for preparing TEM in situ mechanical testing specimens from bulk materials requiring depth-specific examination, e.g., target dose on the ion irradiation damage profile. This study demonstrates a new method for extracting depth-specific TEM in situ tensile bars from a bulk specimen of Fe-5 wt.%Mo. Measured yield stress, ultimate tensile stress, Young’s modulus, and elongation are consistent with those properties obtained from similarly sized Fe and Mo single-crystal nanowires. Results are discussed in the context of the specimen size effect.
"Method for fabricating depth-specific TEM in situ tensile bars" Janelle Wharry, George Warren, Patrick Warren, Megha Dubey, Jatuporn Burns, Yaqiao Wu, JOM Vol. 72 2020 2057-2064 Link
"Microstructure of laser weld repairs of irradiated austenitic steels" Janelle Wharry, Keyou Mao, Aaron French, Xiang Liu, Yaqiao Wu, Cheng Sun, Paula Freyer, Jonathan Tatman, Lucille Giannuzzi, Frank Garner, Lin Shao, Materials & Design Vol. 206 2021 109764 Link
"Microstructure-property relationships across AISI 304/308L stainless steel laser weldments" Janelle Wharry, Keyou Mao, Vikas Tomar, Yaqiao Wu, Materials Science & Engineering A Vol. 721 2018 234-243 Link
"Precession electron diffraction for SiC grain boundary characterization in unirradiated TRISO fuel" Thomas Lillo, Isabella van Rooyen, Yaqiao Wu, Nuclear Engineering and Design Vol. 305 2016 277-283 Link
Precession electron diffraction (PED), a transmission electron microscopy-based technique, has been evaluated for the suitability for evaluating grain boundary character in the SiC layer of tristructural isotropic (TRISO) fuel. This work reports the effect of transmission electron microscope (TEM) lamella thickness on the quality of data and establishes a baseline comparison to SiC grain boundary characteristics, in an unirradiated TRISO particle, determined previously using a conventional electron backscatter diffraction (EBSD) scanning electron microscope (SEM)-based technique. In general, it was determined that the lamella thickness produced using the standard focused ion beam (FIB) fabrication process (∼80 nm), is sufficient to provide reliable PED measurements, although thicker lamellae (∼120 nm) were found to produce higher quality orientation data. Also, analysis of SiC grain boundary character from the TEM-based PED data showed a much lower fraction of low-angle grain boundaries compared to SEM-based EBSD data from the SiC layer of a TRISO-coated particle made using the same fabrication parameters and a SiC layer deposited at a slightly lower temperature from a surrogate TRISO particle. However, the fractions of high-angle and coincident site lattice (CSL)-related grain boundaries determined by PED are similar to those found using SEM-based EBSD. Since the grain size of the SiC layer of TRSIO fuel can be as small as 250 nm (Kirchhofer et al., 2013), depending on the fabrication parameters, and since grain boundary fission product precipitates in irradiated TRISO fuel can be nano-sized, the TEM-based PED orientation data collection method is preferred to determine an accurate representation of the relative fractions of low-angle, high-angle, and CSL-related grain boundaries. It was concluded that although the resolution of the PED data is better by more than an order of magnitude, data acquisition times may be significantly longer or the number of areas analyzed needs to be significantly greater than the SEM-based method to obtain a statistically relevant distribution. Also, grain size could be accurately determined but significantly larger analysis areas would be required than those used in this study.
"Processing of a novel nanostructured ferritic steel via spark plasma sintering and investigation of its mechanical and microstructural characteristics" Darryl Butt, Indrajit Charit, James Cole, Somayeh Pasebani, Yaqiao Wu, Jatuporn Burns, Kerry Allahar, INIS Repository Vol. 46 2015 Link
Nano-structured ferritic steels (NFSs) with 12-14 wt% Cr have attracted widespread interest for potential high temperature structural and fuel cladding applications in advanced nuclear reactors. They have excellent high temperature mechanical properties and high resistance to radiation-induced damage. The properties of the NFSs depend on the composition that mainly consists of Cr, Ti, W or Mo, and Y2O3 as alloying constituents. In this study, a novel nano-structured ferritic steel (Fe-14Cr-1Ti-0.3Mo-0.5La2O3, wt%) termed as 14LMT was developed via high energy ball milling and spark plasma sintering. Vickers microhardness values were measured. Microstructural studies of the developed NFSs were performed by EBSD and TEM, which revealed a bimodal grain size distribution. A significant number density of nano-precipitates was observed in the microstructure. The diameter of the precipitates varied between 2-70 nm and the morphology from the spherical to faceted shape. The Cr-La-Ti-O-enriched nano-clusters were identified by APT studies.
"Sintering Behavior of Lanthana-Bearing Nanostructured Ferritic Steel Consolidated via Spark Plasma Sintering" Darryl Butt, Indrajit Charit, James Cole, Somayeh Pasebani, Yaqiao Wu, Jatuporn Burns, Advanced Engineering Materials Vol. 18 2015 324-332 Link
Elemental powder mixture of Fe–14Cr–1Ti–0.3Mo–0.5La2O3(wt%) composition is mechanicallyalloyed for different milling durations (5, 10 and 20 h) and subsequently consolidated via spark plasmasintering under vacuum at 950?C for 7 min. The effects of milling time on the densi?cation behaviorand density/microhardness are studied. The sintering activation energy is found to be close to that ofgrain boundary diffusion. The bimodal grain structure created in the milled and sintered material isfound to be a result of milling and not of sintering alone. The oxide particle diameter varies between2 and 70 nm. Faceted precipitates smaller than 10 nm in diameter are found to be mostly La–Ti–Cr-enriched complex oxides that restrict further recrystallization and related phenomena
"Stability of nanoclusters in an oxide dispersion strengthened alloy under neutron irradiation" Xiang Liu, Yinbin Miao, Yaqiao Wu, Stuart Maloy, James Stubbins, Scripta Materialia Vol. 138 2017 57-61 Link
Here, we report atom probe tomography results of the nanoclusters in a neutron-irradiated oxide dispersion strengthened alloy. Following irradiation to 5 dpa at target temperatures of 300 °C and 450 °C, fewer large nanoclusters were found and the residual nanoclusters tend to reach an equilibrium Guinier radius of 1.8 nm. With increasing dose, evident decrease in peak oxygen and titanium (but not yttrium) concentrations in the nanoclusters was observed, which was explained by atomic weight, solubility, diffusivity, and chemical bonding arguments. The chemical modifications indicate the equilibrium size is indeed a balance of two competing processes: radiation enhanced diffusion and collisional dissolution.
"STEM-EDS Analysis of Fission Products in Neutron-Irradated TRISO Fuel Particles from AGR-1 Experiment" bin leng, Kumar Sridharan, Izabela Szlufarska, Yaqiao Wu, Isabella van Rooyen, Journal of Nuclear Materials Vol. 475 2016 62-70 Link
Historic and recent post-irradiation-examination from the German AVR and Advanced Gas Reactor Fuel Development and Qualification Project have shown that 110 m Ag is released from intact tristructural isotropic (TRISO) fuel. Although TRISO fuel particle research has been performed over the last few decades, little is known about how metallic fission products are transported through the SiC layer, and it was not until March 2013 that Ag was first identified in the SiC layer of a neutron-irradiated TRISO fuel particle. The existence of Pd- and Ag-rich grain boundary precipitates, triple junction precipitates, and Pd nano-sized intragranular precipitates in neutron-irradiated TRISO particle coatings was investigated using Scanning Transmission Electron Microscopy and Energy Dispersive Spectroscopy analysis to obtain more information on the chemical composition of the fission product precipitates. A U-rich fission product honeycomb shape precipitate network was found near a micron-sized precipitate in a SiC grain about ∼5 μm from the SiC-inner pyrolytic carbon interlayer, indicating a possible intragranular transport path for uranium. A single Ag-Pd nano-sized precipitate was found inside a SiC grain, and this is the first research showing such finding in irradiated SiC. This finding may possibly suggest a possible Pd-assisted intragranular transport mechanism for Ag and may be related to void or dislocation networks inside SiC grains. Preliminary semi-quantitative analysis indicated the micron-sized precipitates to be Pd2Si2U with carbon existing inside these precipitates. However, the results of such analysis for nano-sized precipitates may be influenced by the SiC matrix. The results reported in this paper confirm the co-existence of Cd with Ag in triple points reported previously.
"TEM in situ micropillar compression tests of ion irradiated oxide dispersion strengthened alloy" Matthew Swenson, Janelle Wharry, Yaqiao Wu, Kayla Yano, Journal of Nuclear Materials Vol. 483 2017 107 Link
The growing role of charged particle irradiation in the evaluation of nuclear reactor candidate materials requires the development of novel methods to assess mechanical properties in near-surface irradiation damage layers just a few micrometers thick. In situ transmission electron microscopic (TEM) mechanical testing is one such promising method. In this work, microcompression pillars are fabricated from a Fe2+ ion irradiated bulk specimen of a model Fe-9%Cr oxide dispersion strengthened (ODS) alloy. Yield strengths measured directly from TEM in situ compression tests are within expected values, and are consistent with predictions based on the irradiated microstructure. Measured elastic modulus values, once adjusted for the amount of deformation and deflection in the base material, are also within the expected range. A pillar size effect is only observed in samples with minimum dimension =100 nm due to the low inter-obstacle spacing in the as received and irradiated material. TEM in situ micropillar compression tests hold great promise for quantitatively determining mechanical properties of shallow ion-irradiated layers.
"The Role of Cr, P, and N solutes on the irradiated microstructure of bcc Fe" Patrick Warren, Caleb Clement, Chao Yang, Amrita Sen, Wei-Ying Chen, Yaqiao Wu, Journal of Nuclear Materials Vol. 583 [unknown] Link
The objective of this study is to understand irradiation-induced and assisted defect evolution in binary body center cubic (bcc) Fe-based alloys. The broader class of bcc ferritic alloys are leading candidates for advanced nuclear fission and fusion applications, in part due to their exceptional void swelling resistance. However, their irradiated microstructure evolution is sensitive to solute species present, since these solutes can act as traps for irradiation-induced defects due to the surrounding tensile or compressive stress fields. Here, three alloys (Fe- 9.5%Cr, Fe-4.5%P, and Fe-2.3%N) are selected for study because they systematically exhibit varying solute sizes and solute positions (i.e., substitutional or interstitial). Ex situ and in situ ion irradiations reveal that Fe-P has a considerably finer and denser population of irradiation-induced defects than Fe-Cr and Fe-N at the same irra- diation conditions, which is attributed to strong defect trapping at undersized substitutional P, consequently hindering the development of extended defects. Meanwhile, oversized substitutional solutes (e.g., Cr) and interstitial solutes (e.g., N) may also suppress dislocation loop development due to weak solute-defect trapping.
"The role of Cr, P, and N solutes on the irradiated microstructure of bcc Fe" Janelle Wharry, Patrick Warren, Caleb Clement, Amrita Sen, Chao Yang, Wei-Ying Chen, Yaqiao Wu, Ling Wang, Journal of Nuclear Materials Vol. 583 2023 154531 Link
"Thermomechanical Properties of Neutron Irradiated Al3Hf-Al Thermal Neutron Absorber Materials" Donna Guillen, Mychailo Toloczko, Ramprashad Prabhakaran, Yuanyuan Zhu, Yu Lu, Yaqiao Wu, Materials Vol. 16 2023 5518 Link
thermal neutron absorber material composed of Al3Hf particles in an aluminum matrix is under development for the Advanced Test Reactor. This metal matrix composite was fabricated via hot pressing of high-purity aluminum and micrometer-size Al3Hf powders at volume fractions of 20.0, 28.4, and 36.5%. Room temperature tensile and hardness testing of unirradiated specimens revealed a linear relationship between volume fraction and strength, while the tensile data showed a strong decrease in elongation between the 20 and 36.5% volume fraction materials. Tensile tests conducted at 200 °C on unirradiated material revealed similar trends. Evaluations were then conducted on specimens irradiated at 66 to 75 °C to four dose levels ranging from approximately 1 to 4 dpa. Tensile properties exhibited the typical increase in strength and decrease in ductility with dose that are common for metallic materials irradiated at ≤0.4Tm. Hardness also increased with neutron dose. The difference in strength between the three different volume fraction materials was roughly constant as the dose increased. Nanoindentation measurements of Al3Hf particles in the 28.4 vol% material showed the expected trend of increased hardness with irradiation dose. Transmission electron microscopy revealed oxygen at the interface between the Al3Hf particles and aluminum matrix in the irradiated material. Scanning electron microscopy of the exterior surface of tensile tested specimens revealed that deformation of the material occurs via plastic deformation of the Al matrix, cracking of the Al3Hf particles, and to a lesser extent, tearing of the matrix away from the particles. The fracture surface of an irradiated 28.4 vol% specimen showed failure by brittle fracture in the particles and ductile tearing of the aluminum matrix with no loss of cohesion between the particles and matrix. The coefficient of thermal expansion decreased upon irradiation, with a maximum change of −6.3% for the annealed irradiated 36.5 vol% specimen.
"Assessing mechanical properties of irradiated materials by nanomechanical testing" Janelle Wharry, Kayla Yano, Yaqiao Wu, ICSMA18 July 15-18, (2018)
"Atom probe analysis of a neutron irradiated Fe-14Cr model alloy" Jian Gan, Stuart Maloy, Yinbin Miao, Kun Mo, Yaqiao Wu, ICFRM 2013 January 1-9, (2013)
"In situ TEM mechanical testing: an emerging approach for characterization of polycrystalline, irradiated alloys" Matthew Swenson, Janelle Wharry, Yaqiao Wu, Kayla Yano, Microscopy & Microanalysis July 24-28, (2016)
"Mechanics of irradiated alloys studied through in situ TEM testing" Janelle Wharry, Kayla Yano, Yaqiao Wu, TMS 2018 March 11-15, (2018)
"Microstructure and Mechanical Property Studies on Neutron-Irradiated Ferritic FeCr Model Alloys" Jian Gan, Stuart Maloy, Yinbin Miao, Kun Mo, James Stubbins, Yaqiao Wu, TMS Annual Meeting February 16-20, (2014)
"Processing of a Novel Nanostructured Ferritic Steel via Spark Plasma Sintering and Investigation of Its Mechanical and Microstructural Characteristics" Darryl Butt, Indrajit Charit, James Cole, Somayeh Pasebani, Yaqiao Wu, SMINS-3 October 7-10, (2013)
"Qualitative Analysis of deformation in Proton Irradiated Nanocrystalline Copper Tantalum Alloy" Priyam Patki, Janelle Wharry, Yaqiao Wu, The Minerals, Metals and Materials Society March 10-14, (2019)
"TEM in situ mechanical testing of proton irradiated nanocrystalline copper tantalum alloy" Yaqiao Wu, Janelle Wharry, TMS 2018 March 11-15, (2018)
"The role of electron and atom probe tomography in characterization of nuclear fuels" Assel Aitkaliyeva, Cynthia Papesch, Yaqiao Wu, Haiming Wen, Nuclear Fuels and Structural Materials (NFSM-2016) June 12-16, (2016)
NSUF Articles:
Users Organization Meeting Presentations Now Available - Wednesday, March 25, 2020 - Newsletter, Users Group
CINR Awards Announced - Eight projects were selected Projects will take advantage of NSUF capabilities to investigate important nuclear fuel and material applications. Thursday, June 27, 2019 - Calls and Awards
NSUF Research Collaborations