Yong Yang

Profile Information

Name: Yong Yang
Institution: University of Florida
Position: Associate Professor
Affiliation: University of Florida
h-Index
ORCID: 0000-0002-0247-6219
Expertise: Radiation Damage

Publications:

		"Characterization of microstructure and property evolution in advanced cladding and duct: Materials exposed to high dose and elevated temperature" Todd Allen, Zhijie Jiao, Djamel Kaoumi, Janelle Wharry, cem topbasi, Aaron Kohnert, Leland Barnard, Alicia Certain, Kevin Field, Gary Was, Dane Morgan, Arthur Motta, Brian Wirth, Yong Yang, Journal of Materials Research Vol. 30 2015 1246-1274 Link Designing materials for performance in high-radiation fields can be accelerated through a carefully chosen combination of advanced multiscale modeling paired with appropriate experimental validation. The studies reported in this work, the combined efforts of six universities working together as the Consortium on Cladding and Structural Materials, use that approach to focus on improving the scientific basis for the response of ferritic–martensitic steels to irradiation. A combination of modern modeling techniques with controlled experimentation has specifically focused on improving the understanding of radiation-induced segregation, precipitate formation and growth under radiation, the stability of oxide nanoclusters, and the development of dislocation networks under radiation. Experimental studies use both model and commercial alloys, irradiated with both ion beams and neutrons. Transmission electron microscopy and atom probe are combined with both first-principles and rate theory approaches to advance the understanding of ferritic–martensitic steels.
		"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.
		"In-situ high-energy X-ray characterization of neutron irradiated HT-UPS stainless steel under tensile deformation" Chi Xu, Xuan Zhang, Yiren Chen, Meimei Li, Jun-Sang Park, Peter Kenesei, Jason Almer, Yong Yang, Acta Materialia Vol. 156 2018 330-341 Link The tensile deformation behavior of a high-temperature, ultrafine-precipitate strengthened (HT-UPS) stainless steel was characterized in-situ with high-energy X-ray diffraction at 20 and 400?°C. The HT-UPS samples were neutron irradiated to 3 dpa at 400?°C. Significant irradiation hardening and ductility loss were observed at both temperatures. Lattice strain evolutions of the irradiated samples showed a strong linear response up to near the onset of the macroscopic yield, in contrast to the unirradiated HT-UPS which showed a pronounced non-linear behavior well below the macroscopic yield. While the room-temperature diffraction elastic moduli in the longitudinal direction increased after irradiation, the 400?°C moduli were similar before and after irradiation. The evolution of the {200} lattice strain parallel to the loading axis () showed unique characteristics: in the plastic regime, the evolution of after yield is temperature-dependent in the unirradiated specimens but temperature-independent in the irradiated specimens; and the value of at the yield is an irradiation-sensitive, temperature-independent parameter. The evolution of corresponds well with the dislocation density evolution, and is an effective probe of the deformation-induced long-range internal stresses in the HT-UPS steel.
		"Irradiaiton Response of the Ferrite Phase in CF3 Cast Stainless Steel" Zhangbo Li, Yong Yang, Transactions of the American Nuclear Society Vol. 116 2017 390-391 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.
		"Micro-mechanical evaluation of SiC-SiC composite interphase properties and debond mechanisms" Mehdi Balooch, Peter Hosemann, Cameron Howard, Yutai Katoh, Takaaki Koyanagi, Yong Yang, Joey Kabel, Kurt Terrani, Composites Part B: Engineering Vol. 131 2017 173-183 Link SiC-SiC composites exhibit exceptional high temperature strength and oxidation properties making them an advantageous choice for accident tolerant nuclear fuel cladding. In the present work, small scale mechanical testing along with AFM and TEM analysis were employed to evaluate PyC interphase properties that play a key role in the overall mechanical behavior of the composite. The Mohr-Coulomb formulation allowed for the extraction of the internal friction coefficient and debonding shear strength as a function of the PyC layer thickness, an additional parameter. These results have led to re-evaluation of the Mohr-Coulomb failure criterion and adjustment via a new phenomenological equation.SiC-SiC composites exhibit exceptional high temperature strength and oxidation properties making them an advantageous choice for accident tolerant nuclear fuel cladding. In the present work, small scale mechanical testing along with AFM and TEM analysis were employed to evaluate PyC interphase properties that play a key role in the overall mechanical behavior of the composite. The Mohr-Coulomb formulation allowed for the extraction of the internal friction coefficient and debonding shear strength as a function of the PyC layer thickness, an additional parameter. These results have led to re-evaluation of the Mohr-Coulomb failure criterion and adjustment via a new phenomenological equation.
		"Microstructural evolution of neutron-irradiated T91 and NF616 to ~4.3 dpa at 469 °C" Kevin Field, Bong Goo Kim, Lizhen Tan, Yong Yang, Sean Gray, Meimei Li, Journal of Nuclear Materials Vol. 493 2017 12-20 Link Ferritic-martensitic steels such as T91 and NF616 are candidate materials for several nuclear applications. This study evaluates radiation resistance of T91 and NF616 by examining their microstructural evolutions and hardening after the samples were irradiated in the Advanced Test Reactor to ∼4.3 displacements per atom (dpa) at an as-run temperature of 469 °C. In general, this irradiation did not result in significant difference in the radiation-induced microstructures between the two steels. Compared to NF616, T91 had a higher number density of dislocation loops and a lower level of radiation-induced segregation, together with a slightly higher radiation-hardening. Unlike dislocation loops developed in both steels, radiation-induced cavities were only observed in T91 but remained small with sub-10 nm sizes. Other than the relatively stable M23C6, a new phase (likely Sigma phase) was observed in T91 and radiation-enhanced MX → Z phase transformation was identified in NF616. Laves phase was not observed in the samples.
		"Microstructural evolution of NF709 (20Cr–25Ni–1.5 MoNbTiN) under neutron irradiation" Bong Goo Kim, Lizhen Tan, Yong Yang, Cheryl Xu, Xuan Zhang, Meimei Li, Journal of Nuclear Materials Vol. 470 2016 229-235 Link Because of its superior creep and corrosion resistance as compared with general austenitic stainless steels, NF709 has emerged as a candidate structural material for advanced nuclear reactors. To obtain fundamental information about the radiation resistance of this material, this study examined the microstructural evolution of NF709 subjected to neutron irradiation to 3 displacements per atom at 500 °C. Transmission electron microscopy, scanning electron microscopy, and high-energy x-ray diffraction were employed to characterize radiation-induced segregation, Frank loops, voids, as well as the formation and reduction of precipitates. Radiation hardening of ∼76% was estimated by nanoindentation, approximately consistent with the calculation according to the dispersed barrier-hardening model, suggesting Frank loops as the primary hardening source.
		"Microstructural evolution of NF709 austenitic stainless steel under in-situ ion irradiations at room temperature, 300, 400, 500 and 600 °C" Chi Xu, Wei-Ying Chen, Yiren Chen, Yong Yang, Journal of Nuclear Materials Vol. 509 2018 644-653 Link
		"Microstructure and Fission Product Distribution Examination in the UCO kernel of TRISO Fuel Particles" Isabella van Rooyen, Yong Yang, Terry Holesinger, Mukesh Bachhav, OSTI.gov, Conf Proceedings Vol. 2018 Link
		"Using a spherical crystallite model with vacancies to relate local atomic structure to irradiation defects in ZrC and ZrN" Todd Allen, Hasitha Ganegoda, Daniel Olive, Jeff Terry, Yong Yang, Clayton Dickerson, Journal of Nuclear Materials Vol. 475 2016 123-131 Link Zirconium carbide and zirconium nitride are candidate materials for new fuel applications due to several favorable physicochemical properties. ZrC and ZrN samples were irradiated at the Advanced Test Reactor National Scientific User Facility with neutrons at 800 °C to a dose of 1 dpa. Structural examinations have been made of the ZrC samples using high resolution transmission electron microscopy, and the findings compared with a previous study of ZrC irradiated with protons at 800 °C. The use of X-ray absorption fine structure spectroscopy (XAFS) to characterize the radiation damage was also explored including a model based on spherical crystallites that can be used to relate EXAFS measurements to microscopy observations. A loss of coordination at more distant coordination shells was observed for both ZrC and ZrN, and a model using small spherical crystallites suggested this technique can be used to study dislocation densities in future studies of irradiated materials.

Presentations:

	"Effect of Ni on Formatoin of Entermetallic Phases in Highly Irradiated Reactor Pressure Vessel Steels" G. Robert Odette, Nicolas Silva, Peter Wells, Yong Yang, TMS 2014 February 16-20, (2014)
	"Irradiation Effects in Aged Cast Duplex Stainless Steels" Janne Pakarinen, Yong Yang, TMS 2014 February 16-20, (2014)
	"Irradiation Response of the Ferrite Phase in CF3 Cast Stainless Steel" Zhangbo Li, Yong Yang, 2017 ANS Annual Meeting [unknown]
	"Microstructural Evolution of NF709 Steel under In-situ Ion Irradiations at Room Temperature to 600 °C" Yong Yang, 2017 ANS Annual Meeting [unknown]
	"Microstructure and Fission Product Distribution Examination in the UCO Kernel of TRISO Fuel Particles" Isabella van Rooyen, Yong Yang, Terry Holesinger, , HTR 2018 October 8-10, (2018)

NSUF Articles:

U.S. DOE Nuclear Science User Facilities Awards 35 Rapid Turnaround Experiment Research Proposals - Awards total approximately $1.3 million These projects will continue to advance the understanding of irradiation effects in nuclear fuels and materials in support of the mission of the DOE Office of Nuclear Energy. Wednesday, September 20, 2017 - Calls and Awards
DOE awards 39 RTE Projects - Projects total approximately $1.3 million Thursday, February 1, 2018 - Calls and Awards
NSUF Researcher Feature: Kumar Sridharan - Learn more about a University of Wisconsin professor who helped kick start NSUF Sridharan's research team put the NSUF's first material samples into the ATR, launching a new era of research into the behaviors of fuels and materials in a nuclear reactor environment. Wednesday, August 28, 2019 - Newsletter, Researcher Highlight
DOE Awards 31 RTE Proposals, Opens FY-20 1st Call - Projects total $1.1 million; Next proposals due 10/31 Awards will go to 22 principal investigators from universities, six from national laboratories, and three from foreign universities. Tuesday, September 17, 2019 - Calls and Awards, Announcement

Accomplishments

Accomplished Author

Authored 10+ NSUF-supported publications

Accomplished Reviewer

Reviewed 10+ RTE Proposals

NSUF Supported Research

Understand the phase transformation of thermal aged and neutron irradiated duplex stainless steels used in LWRs - FY 2016 CINR, #16-10696

Understand the Fission Products Behavior in UCO Fuel Kernels of safety tested AGR2 TRISO Fuel Particles by Using Titan Themis 200 with ChemiSTEM Capability - FY 2019 RTE 3rd Call, #19-2893

Understand the Fission Products Behavior and Irradiation Effects in UCO Fuel Kernels of Irradiated AGR-1 and AGR-2 TRISO Fuel Particles Using Titan Themis 200 with ChemiSTEM Capability - FY 2018 RTE 1st Call, #18-1257

Characterize the Irradiated Microstructure and Understand the Fission Product Behavior in an Irradiated and Safety Tested AGR-1 TRISO Fuel Particle New Proposal - FY 2017 RTE 3rd Call, #17-1091

Low temperature Fe-ion irradiation of 15-15Ti steel in different thermo-mechanical states - FY 2017 RTE 1st Call, #16-864

Evaluate the ferrite decomposition in irradiated duplex cast stainless steels - FY 2016 RTE 2nd Call, #16-655

Characterize Neutron Irradiated NF709 Stainless Steel Using Atom Probe Tomography - FY 2015 RTE 3rd Call, #15-591

Characterization on the Bor-60 neutron irradiated austenitic stainless steels and cast stainless steel - FY 2015 RTE 2nd Call, #15-564

Synergistic Effects of Thermal aging and Neutron Irradiation in 304L Welds - FY 2015 RTE 1st Call, #15-531

Irradiation Effects inAged Cast Duplex Stainless Steels - FY 2013 RTE Solicitation, #13-411

Radiation Stability of Ceramics for Advanced Fuel Applications - FY 2009 Fall Solicitation for User Proposals, #09-152

NSUF Research Collaborations

Understand the Fission Products Behavior and Irradiation Effects in UCO Fuel Kernels of Irradiated AGR-1 and AGR-2 TRISO Fuel Particles by Using Atom Probe Tomography - FY 2020 RTE 2nd Call, #20-4217

Predict the mechanical behavior of irradiated cast stainless steels based on the microstructures and measured properties from nanoindentation - FY 2019 RTE 3rd Call, #19-2895

Understand the atomic positions of the metallic fission product in UCO Fuel Kernels and Determine the exact stoichiometry of UC, UO phase of Irradiated TRISO Fuel Particles by Using Titan Themis 200 with EELS characterization Capability - FY 2019 RTE 2nd Call, #19-1779

Microstructure characterization on neutron irradiated and post-tensile duplex stainless steels - FY 2019 RTE 1st Call, #19-1672

Understand the Fission Products Behavior and Irradiation Effects in UCO Fuel Kernels of Irradiated AGR-1 and AGR-2 TRISO Fuel Particles by Using Atom Probe Tomography - FY 2018 RTE 3rd Call, #18-1593

Fission Product Distribution Comparison in Irradiated and Safety Tested AGR-1 and AGR-2 TRISO Fuel Particles - FY 2018 RTE 1st Call, #18-1256

An in-situ TEM characterization of tensile testing of ion irradiated HT-UPS steel at RT and 400°C - FY 2017 RTE 3rd Call, #17-1098

Characterization of ion irradiated 15-15Ti steel by APT - FY 2017 RTE 2nd Call, #17-930

Irradiation of the vanadium carbide coating on HT-9 steel using protons - FY 2017 RTE 1st Call, #16-863

Effect of Neutron Irradiation on Tensile Deformation of a HT-UPS Stainless Steel - FY 2016 RTE 2nd Call, #16-643

Irradiation Effect in the Heterogeneous Hardening of Cast Austenitic Stainless Steels - FY 2016 RTE 2nd Call, #16-649

Characterize Neutron Irradiated HT-UPS Stainless Steel Using Transmission Electron Microscopy and Atom Probe Tomography - FY 2016 RTE 1st Call, #16-622

Critical evaluation of radiation tolerance of nanocrystalline austenitic stainless steels - FY 2012 RTE Solicitation, #12-353

Yong Yang

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