James Stubbins

Profile Information

Name: Professor James Stubbins
Institution: University of Illinois
Position: Willett Professor
h-Index
ORCID: 0000-0003-4559-5848

Publications:

		"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.
		"Early-stage microstructural evolution and phase stability in neutron-irradiated ferritic-martensitic steel T91" Huan Yan, Xiang Liu, Lingfeng He, James Stubbins, Journal of Nuclear Materials Vol. 557 2021 Link
		"High energy X-ray diffraction study of the relationship between the macroscopic mechanical properties and microstructure of irradiated HT-9 steel" Osman Anderoglu, Bjorn Clausen, James Stubbins, Journal of Nuclear Materials Vol. 475 2016 46-56 Link
		"In situ high-energy X-ray diffraction study of tensile deformation of neutron-irradiated polycrystalline Fe-9%Cr alloy" Meimei Li, James Stubbins, Chi Xu, Xuan Zhang, Jun-Sang Park, Peter Kenesei, Jonathan Almer, Acta Materialia Vol. 126 2017 67-76 Link The effect of neutron irradiation on tensile deformation of a Fe-9wt%Cr alloy was investigated using in situ high-energy synchrotron X-ray diffraction during room-temperature uniaxial tension tests. New insights into the deformation mechanisms were obtained through the measurements of lattice strain evolution and the analysis of diffraction peak broadening using the modified Williamson-Hall method. Two neutron-irradiated specimens, one irradiated at 300 °C to 0.01 dpa and the other at 450 °C to 0.01 dpa, were tested along with an unirradiated specimen. The macroscopic stress–strain curves of the irradiated specimens showed increased strength, reduced ductility and reduced work-hardening exponent compared to the unirradiated specimen. The evolutions of the lattice strain, the dislocation density and the coherent scattering domain size in the deformation process revealed different roles of the submicroscopic defects in the 300 °C/0.01 dpa specimen and the nanometer-sized dislocation loops in the 450 °C/0.01 dpa specimen; the dislocation loops were more effective in dislocation pinning. While the work hardening rate of stage II was unaffected by irradiation, significant dynamic recovery in stage III in the irradiated specimens led to the early onset of necking without stage IV as observed in the unirradiated specimen.
		"Investigation of thermal aging effects on the tensile properties of Alloy 617 by in-situ synchrotron wide-angle X-ray scattering" Xiang Liu, Kun Mo, Yinbin Miao, Kuan-Che Lan, Guangming Zhang, Wei-Ying Chen, Carolyn Tomchik, Rachel Seibert, Jeff Terry, James Stubbins, Materials Science and Engineering: A Vol. 651 2016 55-62 Link The nickel-base Alloy 617 has been considered as the lead candidate structural material for the intermediate heat exchanger (IHX) of the Very-High-Temperature Reactor (VHTR). In order to assess the long-term performance of Alloy 617, thermal aging experiments up to 10,000 h in duration were performed at 1000 °C. Subsequently, in-situ synchrotron wide-angle X-ray scattering (WAXS) tensile tests were carried out at ambient temperature. M23C6 carbides were identified as the primary precipitates, while a smaller amount of M6C was also observed. The aging effects were quantified in several aspects: (1) macroscopic tensile properties, (2) volume fraction of the M23C6 phase, (3) the lattice strain evolution of both the matrix and the M23C6 precipitates, and (4) the dislocation density evolution during plastic deformation. The property?microstructure relationship is described with a focus on the evolution of the M23C6 phase. For aging up to 3000 h, the yield strength (YS) and ultimate tensile strength (UTS) showed little variation, with average values being 454 MPa and 787 MPa, respectively. At 10,000 h, the YS and UTS reduced to 380 MPa and 720 MPa, respectively. The reduction in YS and UTS is mainly due to the coarsening of the M23C6 precipitates. After long term aging, the volume fraction of the M23C6 phase reached a plateau and its maximum internal stress was reduced, implying that under large internal stresses the carbides were more susceptible to fracture or decohesion from the matrix. Finally, the calculated dislocation densities were in good agreement with transmission electron microscopy (TEM) measurements. The square roots of the dislocation densities and the true stresses displayed typical linear behavior and no significant change was observed in the alloys in different aging conditions.
		"Ion-irradiation-induced microstructural modifications in ferritic/martensitic steel T91" Xiang Liu, Yinbin Miao, Meimei Li, Marquis Kirk, Stuart Maloy, James Stubbins, Journal of Nuclear Materials Vol. 490 2017 305-316 Link In this paper, in situ transmission electron microscopy investigations were carried out to study the microstructural evolution of ferritic/martensitic steel T91 under 1 MeV Krypton ion irradiation up to 4.2 × 1015 ions/cm2 at 573 K, 673 K, and 773 K. At 573 K, grown-in defects are strongly modified by black-dot loops, and dislocation networks together with black-dot loops were observed after irradiation. At 673 K and 773 K, grown-in defects are only partially modified by dislocation loops; isolated loops and dislocation segments were commonly found after irradiation. Post irradiation examination indicates that at 4.2 × 1015 ions/cm2, about 51% of the loops were type for the 673 K irradiation, and the dominant loop type was for the 773 K irradiation. Finally, a dispersed barrier hardening model was employed to estimate the change in yield strength, and the calculated ion data were found to follow the similar trend as the existing neutron data with an offset of 100–150 MPa.
		"Kinetics of the Migration and Clustering of Extrinsic Gas in bcc Metals" Maria Okuniewski, James Stubbins, Chaitanya Deo, Srinivasan Srivilliputhur, Michael Baskes, Stuart Maloy, M. R. James, ASTM Special Technical Publication Vol. 1492 2008 177-189 Link We study the mechanisms by which gas atoms such as helium and hydrogen diffuse and interact with other defects in bcc metals and investigate the effect of these mechanisms on the nucleation of embryonic gas bubbles. Large quantities of helium and hydrogen are produced due to spallation and transmutation in structural materials in fusion and accelerator-driven reactors. The long time evolution of the extrinsic gas atoms and their accumulation at vacancies is studied using a kinetic Monte Carlo algorithm that is parameterized by the migration energies of the point defect entities. First-order reaction kinetics are observed when gas clusters with vacancies. If gas-gas clustering is allowed, mixed-order diffusion limited kinetics are observed. When dissociation of gas from clusters is allowed, gas-vacancy clusters survive to steady state while gas-gas clusters dissolve. We obtain cluster size distributions and reaction rate constants that can be used to quantify microstructural evolution of the irradiated metal.
		"Lattice strain and damage evolution of 9-12%Cr ferritic/martensitic steel during in situ tensile test by X-ray diffraction and small angle scattering" Kun Mo, James Stubbins, Xiao Pan, Xianglin Wu, Xiang Chen, Jonathan Almer, Jan Ilavsky, Dean Haeffner, Journal of Nuclear Materials Vol. 407 2010 10-15 Link In situ X-ray diffraction and small angle scattering measurements during tensile tests were performed on 9–12% Cr ferritic/martensitic steels. The lattice strains in both particle and matrix phases, along two principal directions, were directly measured. The load transfer between particle and matrix was calculated based on matrix/particle elastic mismatch, matrix plasticity and interface decohesion. In addition, the void or damage evolution during the test was measured using small angle X-ray scattering. By combining stress and void evolution during deformation, the critical interfacial strength for void nucleation was determined, and compared with pre-existing void nucleation criteria. These comparisons show that models overestimate the measured critical strength, and require a larger particle size than measured to match the X-ray observations.
		"Neutron irradiation effects in Fe and Fe-Cr at 300°C" Wei-Ying Chen, Yinbin Miao, Jian Gan, Maria Okuniewski, Stuart Maloy, James Stubbins, Acta Materialia Vol. 111 2016 407-416 Link Fe and Fe-Cr (Cr = 10-16 at.%) specimens were neutron-irradiated at 300 °C to 0.01, 0.1 and 1 dpa. The TEM observations indicated that the Cr significantly reduced the mobility of dislocation loops and suppressed vacancy clustering, leading to distinct damage microstructures between Fe and Fe-Cr. Irradiation-induced dislocation loops in Fe were heterogeneously observed in the vicinity of grown-in dislocations, whereas the loop distribution observed in Fe-Cr is much more uniform. Voids were observed in the irradiated Fe samples, but not in irradiated Fe-Cr samples. Increasing Cr content in Fe-Cr results in a higher density, and a smaller size of irradiation-induced dislocation loops. Orowan mechanism was used to correlate the observed microstructure and hardening, which showed that the hardening in Fe-Cr can be attributed to the formation of dislocation loops and α′ precipitates.
		"Phase stability and microstructural evolution in neutron-irradiated ferritic-martensitic steel HT9" Huan Yan, Xiang Liu, James Stubbins, Journal of Nuclear Materials Vol. 557 2021 Link
		"Radiation resistance of oxide dispersion strengthened alloys: Perspectives from in situ observations and rate theory calculations" Xiang Liu, Yinbin Miao, Meimei Li, Marquis Kirk, Guangming Zhang, Shigeharu Ukai, Stuart Maloy, James Stubbins, Scripta Materialia Vol. 148 2018 33-36 Link Here, in situ ion irradiation and rate theory calculations were employed to directly compare the radiation resistance of an oxide dispersion strengthened alloy with that of a conventional ferritic/martensitic alloy. Compared to the rapid buildup of dislocation loops, loop growth, and formation of network dislocations in the conventional ferritic/martensitic alloy, the superior radiation resistance of the oxide dispersion strengthened alloy is manifested by its stable dislocation structure under the same irradiation conditions. The results are consistent with rate theory calculations, which show that high-density nanoparticles can significantly reduce freely migrating defects and suppress the buildup of clustered defects.
		"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.
		"Temperature and particle size effects on flow localization of 9-12%Cr ferritic/martensitic steel by in situ X-ray diffraction and small angle scattering" Kun Mo, James Stubbins, Xiao Pan, Xianglin Wu, Xiang Chen, Jonathan Almer, Dean Haeffner, Journal of Nuclear Materials Vol. 398 2010 220-226 Link Radiation-induced defect structures are known to elevate material yield strength and reduce material ductility so that small strains induce plastic instability. This process is commonly known as flow localization. Recent research indicates that the flow localization in face-centered cubic (FCC) materials is controlled by critical stress, the true stress at the onset of necking. Critical stress is found to be independent of irradiation dose, but have strong temperature dependence. Here simplified 9–12% ferritic/martinsetic steels are examined using X-ray diffraction and small angle scattering under in situ tensile deformation, in order to elucidate the controlling mechanisms and temperature dependence of critical stress. It is found that the critical stress for the onset of necking is linearly correlated with critical interfacial strength, which in turn determines the void nucleation. The effects of temperature and particle size on critical stress are correspondingly determined by how temperature and particle size influence the critical interfacial strength.
		"The comparison of microstructures and mechanical properties between 14 cr-Al and 14Cr-Ti ferritic ODS alloys" Yinbin Miao, Kun Mo, James Stubbins, Guangming Zhang, Zhangjian Zhou, Shaofu Li, Xiang Liu, Man Wang, Jun-Sang Park, Jonathan Almer, Materials & Design Vol. 98 2016 61-67 Link In this study, two kinds of 14Cr ODS alloys (14Cr-Al and 14Cr-Ti) were investigated to reveal the different effects between Al and Ti on the microstructures and mechanical properties of 14Cr ferritic ODS alloys. The microstructure information such as grains, minor phases of these two alloys has been investigated by high-energy X-ray diffraction and transmission electron microscopy (TEM). The in situ synchrotron X-ray diffraction tensile test was applied to investigate the mechanical properties of these two alloys. The lattice strains of different phases through the entire tensile deformation process in these two alloys were analyzed to calculate their elastic stresses. From the comparison of elastic stress, the strengthening capability of Y2Ti2O7 is better than TiN in 14Cr-Ti, and the strengthening capability of YAH is much better than YAM and AlN in 14Cr-Al ODS. The dislocation densities of 14Cr-Ti and 14Cr-Al ODS alloys during tensile deformation were also examined by modified Williamson-Hall analyses of peak broadening, respectively. The different increasing speed of dislocation density with plastic deformation reveals the better strengthening effect of Y-Ti-O particles in 14Cr-Ti ODS than that of Y-Al-O particles in 14Cr-Al ODS alloy.

Presentations:

	"Ferritic/Martensitic Steels for Fast Reactor Applications" Yinbin Miao, James Stubbins, Huan Yan, International Workshop on Fast Reactor Metallic Fuels RD Program October 10-11, (2017)
	"In-Situ Synchrotron X-Ray Scattering Study on the Tensile Properties of Neutron Irradiated Ferritic/ Martensitic Alloys" Xiang Liu, Kuan-Che Lan, Meimei Li, Xuan Zhang, Chi Xu, James Stubbins, ANS Annual Meeting 2018 June 11-22, (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)
	"Neutron and Ion Irradiation Studies on Advanced Steels Using the Nuclear Science User Facilities" Wei-Ying Chen, Yinbin Miao, James Stubbins, Transactions of the American Nuclear Society June 11-15, (2017)
	"Neutron irradiation damage in ferritic ODS steel MA957" Yinbin Miao, James Stubbins, TMS 2017 146th Annual Meeting & Exhibition February 26-2, (2017)
	"Neutron Radiation Response of Ferritic ODS Steel MA957" Yinbin Miao, James Stubbins, International Conference on Fusion Reactor Materials ICFRM-18 November 5-10, (2017)

NSUF Articles:

NSUF awards 28 Rapid Turnaround Experiment proposals - Approximately $1.74M has been awarded. The new call closes June 28. Thursday, June 1, 2023 - Calls and Awards

Accomplishments

NSUF Profile

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Accomplished Author

Authored 10+ NSUF-supported publications

NSUF Presenter

Presented an NSUF-supported publication

Top 5% Submitter

Top 5% of all RTE Proposal submissions

Top 5% Researcher

Top 5% of all RTE Proposals awarded

Accomplished Collaborator

Collaborated on 3+ RTE Proposals

Accomplished Reviewer

Reviewed 10+ RTE Proposals

NSUF Supported Research

A comparative study of the radiation response of Fe–12Cr, Fe–14Cr, Fe–19Cr model alloys and a Fe–14Cr ODS alloy - FY 2018 RTE 2nd Call, #1433

Advanced Characterizations of Low-dose Neutron Irradiated T91 and HT9 Alloys - FY 2015 CINR, #1711

Characterization of Irradiation-Assisted Stress Corrosion Cracking in 316 Stainless Steel Baffle-Former Bolts Harvested from Commercial Pressurized Water Reactor - FY 2024 CINR, #5023

In-situ Synchrotron Wide-Angle X-ray Scattering (WAXS) Tensile Investigation of Neutron Irradiated Ferritic Alloys - FY 2015 CINR, #1710

Irradiation Performance of Fe-Cr Base Alloys - FY 2008 Call for User Proposals, #92

Microstructure characterization of neutron-irradiated Fe-Cr-C model alloys - FY 2018 RTE 1st Call, #1203

Microstructure characterization of neutron-irradiated Fe-Cr-C model alloys - FY 2019 RTE 2nd Call, #1802

Nanoindentation investigations of neutron-irradiated Fe-Cr-C ternary model alloys - FY 2016 RTE 3rd Call, #694

Positron Annihilation Studies of Neutron Irradiated Ferritic Alloys - FY 2017 CINR, #3058

Post Irradiation Tensile Performance of Fe-Cr Base Alloys - FY 2012 APS, #355

Room Temperature Tensile Properties of ATR Neutron Irradiated T91 - FY 2023 RTE 2nd Call, #4705

TEM determination of dislocation structure formation at 4% tensile deformation in a neutron irradiated Fe-9Cr Model Alloy - FY 2024 RTE 3rd Call, #5157

NSUF Research Collaborations

A study of the tensile response of HT-9 alloys following ATR irradiation to doses between 0.01 and 10 dpa at 300, 450 and 550C - FY 2019 RTE 1st Call, #1670

Atom Probe Tomography Study of Elemental Segregation and Precipitation in Ion-Irradiated Advanced Austenitic Alloy A709 - FY 2020 RTE 1st Call, #2963

Post Irradiation Examination of ATR-irradiated ECAP'ed Steel. - FY 2013 RTE Solicitation, #405

Post-test tensile fractography and microstructure of HT-9 alloys following ATR irradiation to doses between 0.01 and 10 dpa at 300, 450 and 550°C - FY 2024 RTE 2nd Call, #4950