Kumar Sridharan

Profile Information

Name: Professor Kumar Sridharan
Institution: University of Wisconsin
Position: Professor
Biography: Prof. Kumar Sridharan is Grainger Professor in the Departments of Nuclear Engineering & Engineering Physics and Materials Science & Enginerring at the University of Wisconsin, Madison. His expertise spans a broad spectrum of areas in the field of materials science and metallurgy, including nuclear reactor materials, corrosion, physical metallurgy, surface modification and coatings processes, ion implantation, plasma-based synthesis and deposition of materials, characterization and testing of materials, intersections of materials and manufacturing, and industrial applications. He has over 350 publications in these areas including eight invited book chapters, journal articles, reviewed conference proceedings, and industry/national laboratory reports. Prof. Sridharan has provided research mentorship to over 100 graduate and undergraduate students, post-doctoral research associates, staff scientists, and incoming faculty. He serves on the editorial committee of the journal, Advanced Materials & Processes which specializes in reports on cutting-edge, applications-driven materials research and associated manufacturing processes. In 2013 he received the University of Wisconsin, Madison's Chancellor's Award for Excellence in Research. In 2016, he received the Faculty Recognition Award given for enriching and inspiring collegiate experience of undergraduate students by Leaders in Engineering Excellence and Diversity in the University of Wisconsin, Madison's College of Engineering. Prof. Sridharan is an elected Fellow of American Society for Materials, Fellow of Institute of Materials, Minerals, and Mining, United Kingdom, and Fellow of American Nuclear Society in recognition for his contributions to the fields of surface modification and coatings technologies, nuclear reactor materials, and education.

Publications:

		"A combined APT and SANS investigation of a' phase precipitation in neutron-irradiated model FeCrAl alloys" Philip Edmondson, Kevin Field, Kumar Sridharan, Kurt Terrani, Samuel A. Briggs, Kenneth Littrell, Yukinori Yamamoto, Richard Howard, Charles Daily, Acta Materialia Vol. 129 2017 217-228 Link
		"A Novel Approach for Manufacturing Oxide Dispersion Strengthened (ODS) Steel Cladding Tubes using Cold Spray Technology" Ben Maier, Mia Lenling, Stuart Maloy, Kumar Sridharan, Nuclear Engineering and Technology Vol. 2019 Link A novel fabrication method of oxide dispersion strengthened (ODS) steel cladding tubes for advanced fast reactors has been investigated using the cold spray powder-based materials deposition process. Cold spraying has the potential advantage for rapidly fabricating ODS cladding tubes in comparison with the conventional multi-step extrusion process. A gas atomized spherical 14YWT (Fe-14%Cr, 3%W, 0.4%Ti, 0.2%Y, 0.01%O) powder was sprayed on a rotating cylindrical 6061-T6 aluminum mandrel using nitrogen as the propellant gas. The powder lacked the oxygen content needed to precipitate the nanoclusters in ODS steel, therefore this work was intended to serve as a proof-of-concept study to demonstrate that free-standing steel cladding tubes with prototypical ODS composition could be manufactured using the cold spray process. The spray process produced an approximately one-millimeter thick, dense 14YWT deposit on the aluminum-alloy tube. After surface polishing of the 14YWT deposit to obtain desired cladding thickness and surface roughness, the aluminum-alloy mandrel was dissolved in an alkaline medium to leave behind a free-standing ODS tube. The as-fabricated cladding tube was annealed at 1000 °C for 1 hour in an argon atmosphere to improve the overall mechanical properties of the cladding.
		"Atomic Resolution Imaging of Black Spot Defects in Ion Irradiated Silicon Carbide" Li He, Hao Jiang, Yizhang Zhai, Cheng Liu, Izabela Szlufarska, Beata Tyburska-Puschel, Kumar Sridharan, Paul Voyles, Microscopy and Microanalysis Vol. 21 2015 1337-1338 Link
		"Complementary Techniques for Quantification of a' Phase Precipitation in Neutron-Irradiated Fe-Cr-Al Model Alloys" Samuel A. Briggs, Philip Edmondson, Kevin Field, Kumar Sridharan, Yukinori Yamamoto, Kenneth Littrell, Charles Daily, Microscopy & Microanalysis Vol. 22 2016 1470-1471 Link
		"Correlative Microscopy of Neutron-Irradiated Materials" Samuel A. Briggs, Kevin Field, Kumar Sridharan, Advanced Materials & Processes Vol. 174 2016 16-21 Link Development of new, radiation-tolerant materials that maintain the structural integrity and safety margins over the course of a nuclear power reactor’s service life requires the ability to predict degradation phenomena.
		"Corrosion of Structural Alloys in High-Temperature Molten Fluoride Salts for Applications in Molten Salt Reactors" Guiqiu Zheng, Kumar Sridharan, JOM Vol. 70 2018 1535-1541 Link
		"Dependencies of a' embrittlement in neutron-irradiated model Fe-Cr-Al alloys" Samuel A. Briggs, Philip Edmondson, Kevin Field, Kumar Sridharan, ANS Transactions Vol. 114 2016 1046-1047 Link
		"Dislocation loop evolution during in-situ ion irradiation of model FeCrAl alloys" Philip Edmondson, Kevin Field, Jack Haley, Steve Roberts, Kumar Sridharan, Samuel A. Briggs, Sergio Lozano-Perez, Acta Materialia Vol. 136 2017 390-401 Link Model FeCrAl alloys of Fe-10%Cr-5%Al, Fe-12%Cr-4.5%Al, Fe-15%Cr-4%Al, and Fe-18%Cr-3%Al (in wt %) were irradiated with 1 MeV Kr++ ions in-situ with transmission electron microscopy to a dose of 2.5 displacements per atom (dpa) at 320 °C. In all cases, the microstructural damage consisted of dislocation loops with ½<111> and <100> Burgers vectors. The proportion of ½<111> dislocation loops varied from ~50% in the Fe-10%Cr-5%Al model alloy and the Fe-18Cr%-3%Al model alloy to a peak of ~80% in the model Fe-15%Cr-4.5%Al alloy. The dislocation loop volume density increased with dose for all alloys and showed signs of approaching an upper limit. The total loop populations at 2.5 dpa had a slight (and possibly insignificant) decline as the chromium content was increased from 10 to 15 wt %, but the Fe-18%Cr-3%Al alloy had a dislocation loop population ~50% smaller than the other model alloys. The largest dislocation loops in each alloy had image sizes of close to 20 nm in the micrographs, and the median diameters for all alloys ranged from 6 to 8 nm. Nature analysis by the inside-outside method indicated most dislocation loops were interstitial type.
		"Effect of carbon ion irradiation on Ag diffusion in SiC" Tyler Gerczak, bin leng, Kumar Sridharan, Izabela Szlufarska, Hyunseok Ko, Jie Deng, Andrew Giordani, Jerry Hunter, Dane Morgan, Journal of Nuclear Materials Vol. 471 2017 220-232 Link Transport of Ag fission product through the silicon-carbide (SiC) diffusion barrier layer in TRISO fuel particles is of considerable interest given the application of this fuel type in high temperature gas-cooled reactor (HTGR) and other future reactor concepts. The reactor experiments indicate that radiation may play an important role in release of Ag; however so far the isolated effect of radiation on Ag diffusion has not been investigated in controlled laboratory experiments. In this study, we investigate the diffusion couples of Ag and polycrystalline 3C–SiC, as well as Ag and single crystalline 4H–SiC samples before and after irradiation with C2+ ions. The diffusion couple samples were exposed to temperatures of 1500 °C, 1535 °C, and 1569 °C, and the ensuing diffusion profiles were analyzed by secondary ion mass spectrometry (SIMS). Diffusion coefficients calculated from these measurements indicate that Ag diffusion was greatly enhanced by carbon irradiation due to a combined effect of radiation damage on diffusion and the presence of grain boundaries in polycrystalline SiC samples.
		"Effects of Al and Ti Additions on Irradiation Behavior of FeMnNiCr Multi-Principal-Element Alloy" Andrew Hoffman, Haiming Wen, Li He, Kumar Sridharan, Matthew Luebbe, Jiaqi Duan, JOM Vol. 72 2020 150-159 Two Co-free multi-principal-element alloys (MPEAs), viz. single-phase face-centered cubic (FCC) Fe30Ni30Mn30Cr10 and (Fe30Ni30Mn30Cr10)94Ti2Al4 (all in atomic percent) with FCC matrix containing Ni-Ti-Al enriched L12 (ordered FCC) secondary phase (γ′), have been developed and investigated. The alloys were ion irradiated at 300°C and 500°C to peak damage of 120 displacements per atom (dpa). Compared with the (Fe30Ni30Mn30Cr10)94Ti2Al4 alloy, in the Fe30Ni30Mn30Cr10 alloy, the dislocation loops were smaller, with a higher number density. The difference in loop size between the two MPEAs was attributed to the addition of Ti to the matrix, which was anticipated to lower the stacking fault energy and stabilize the faulted Frank loops. The γ′ phase showed good stability under irradiation, with no new γ′ precipitation or growth in existing precipitates. Both alloys showed similar irradiation-induced hardening at 300°C, but the (Fe30Ni30Mn30Cr10)94Ti2Al4 alloy exhibited lower irradiation-induced hardening at 500°C compared with the Fe30Ni30Mn30Cr10 alloy.
		"Enhanced diffusion of Cr in 20Cr-25Ni type alloys under proton irradiation at 670 °C" Tianyi Chen, Ying Yang, Li He, Beata Tyburska-Puschel, Kumar Sridharan, Haixuan Xu, Lizhen Tan, Nuclear Materials and Energy Vol. 17 2018 142-146 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.
		"Evolution of small defect clusters in ion-irradiated 3C-SiC: Combined cluster dynamics modeling and experimental study" Cheng Liu, Li He, Yizhang Zhai, Beata Tyburska-Puschel, Paul Voyles, Kumar Sridharan, Dane Morgan, Izabela Szlufarska, Acta Materialia Vol. 125 2017 377-389 Link
		"Examining the influence of stacking fault width on deformation twinning in an austenitic stainless steel" Gabriel Meric, Maxim Gussev, Kumar Sridharan, Scripta Materialia Vol. 157 2018 162-166 Link
		"Feasibility Study of Making Metallic Hybrid Materials Using Additive Manufacturing" Kumar Sridharan, Metallurgical and Materials Transactions A Vol. 2018 5035-5041 Link
		"High-Resolution Scanning Transmission Electron Microscopy Study of Black Spot Defects in Ion Irradiated Silicon Carbide" Li He, Yizhang Zhai, Cheng Liu, Chao Jiang, Izabela Szlufarska, Beata Tyburska-Puschel, Kumar Sridharan, Paul Voyles, Microscopy and Microanalysis Vol. 20 2014 1824-1825 Link
		"In situ microstructural evolution in face-centered and body-centered cubic complex concentrated solid-solution alloys under heavy ion irradiation" Michael Moorehead, Calvin Parkin, Mohamed Elbakhshwan, Jing Hu, Wei-Ying Chen, Meimei Li, Lingfeng He, Kumar Sridharan, Adrien Couet, Acta Materialia Vol. 198 2020 85-99 Link This study characterizes the microstructural evolution of single-phase complex concentrated solid- solution alloy (CSA) compositions under heavy ion irradiation with the goal of evaluating mecha- nisms for CSA radiation tolerance in advanced fission systems. Three such alloys, Cr 18 Fe 27 Mn 27 Ni 28 , Cr 15 Fe 35 Mn 15 Ni 35 , and equimolar NbTaTiV, along with reference materials (pure Ni and E90 for the Cr- FeMnNi family and pure V for NbTaTiV) were irradiated at 50 K and 773 K with 1 MeV Kr ++ ions to vari- ous levels of displacements per atom (dpa) using in-situ transmission electron microscopy. Cryogenic irra- diation resulted in small defect clusters and faulted dislocation loops as large as 12 nm in face-centered cubic (FCC) CSAs. With thermal diffusion suppressed at cryogenic temperatures, defect densities were lower in all CSAs than in their less compositionally complex reference materials indicating that point defect production is reduced during the displacement cascade stage. High temperature irradiation of the two FCC CSA resulted in the formation of interstitial dislocation loops which by 2 dpa grew to an average size of 27 nm in Cr 18 Fe 27 Mn 27 Ni 28 and 10 nm in Cr 15 Fe 35 Mn 15 Ni 35 . This difference in loop growth kinet- ics was attributed to the difference in Mn-content due to its effect on the nucleation rate by increasing vacancy mobility or reducing the stacking-fault energy.#171118
		"Irradiation Test Plan for the ATR National Scientific User Facility - University of Wisconsin Pilot Project" Kumar Sridharan, Heather MacLean, Timothy Hyde, INL/EXT-09-15627 Vol. 2008 Link The performance of advanced nuclear systems critically relies on the performance of the materials used for cladding, duct, and other structural components. In many proposed advanced systems, the reactor design pushes the temperature and the total radiation dose higher than typically seen in a light water reactor. Understanding the stability of these materials under radiation is critical. There are a large number of materials or material systems that have been developed for greater high temperature or high dose performance for which little or no information on radiation response exists. The goal of this experiment is to provide initial data on the radiation response of these materials. The objective of the UW experiment is to irradiate materials of interest for advanced reactor applications at a variety of temperatures (nominally 300°C, 400°C, 500°C, and 700°C) and total dose accumulations (nominally 3 dpa and 6 dpa). Insertion of this irradiation test is proposed for September 2008 (ATR Cycle 143A). Table 1. UW Experiment Summary UW Experiment Designation Capsule ID Material ATR Insertion Target Dose UW-0 N/A September 2008 N/A UW-1, UW-2 Metals, Ceramics September 2008 3 dpa UW-3 Metals, Ceramics September 2008 6 dpa
		"Irradiation-enhanced a' precipitation in model FeCrAl alloys" Philip Edmondson, Kevin Field, Kumar Sridharan, Samuel A. Briggs, Yukinori Yamamoto, Richard Howard, Kurt Terrani, Scripta Materialia Vol. 116 2016 112-116 Link Model FeCrAl alloys with varying compositions (Fe(10–18)Cr(10–6)Al at.%) have been neutron irradiated at ~ 320 to damage levels of ~ 7 displacements per atom (dpa) to investigate the compositional influence on the formation of irradiation-induced Cr-rich a' precipitates using atom probe tomography. In all alloys, significant number densities of these precipitates were observed. Cluster compositions were investigated and it was found that the average cluster Cr content ranged between 51.1 and 62.5 at.% dependent on initial compositions. This is significantly lower than the Cr-content of a' in binary FeCr alloys. Significant partitioning of the Al from the a' precipitates was also observed.
		"Measurement of Irradiation-induced Swelling in Stainless Steels with a New Transmission Electron Microscopy Method" Li He, Haixuan Xu, Lizhen Tan, Paul Voyles, Kumar Sridharan, Microscopy and Microanalysis Vol. 23 2017 2234-2235 Link
		"Microstructural evolution in Fe-20Cr-25Ni austenitic alloys under proton irradiation at 670 ºC" Tianyi Chen, Lizhen Tan, Li He, Beata Tyburska-Puschel, Kumar Sridharan, Transactions of American Nuclear Society Vol. 117 2017 581-583 Link
		"Observations of Ag diffusion in ion implanted SiC" Todd Allen, Tyler Gerczak, bin leng, Kumar Sridharan, Jerry Hunter, Andrew Giordani, Journal of Nuclear Materials Vol. 461 2015 314-324 Link The nature and magnitude of Ag diffusion in SiC has been a topic of interest in connection with the performance of tristructural isotropic (TRISO) coated particle fuel for high temperature gas-cooled nuclear reactors. Ion implantation diffusion couples have been revisited to continue developing a more complete understanding of Ag fission product diffusion in SiC. Ion implantation diffusion couples fabricated from single crystal 4H-SiC and polycrystalline 3C-SiC substrates and exposed to 1500–1625 °C, were investigated by transmission electron microscopy and secondary ion mass spectrometry (SIMS). The high dynamic range of SIMS allowed for multiple diffusion régimes to be investigated, including enhanced diffusion by implantation-induced defects and grain boundary (GB) diffusion in undamaged SiC. Estimated diffusion coefficients suggest GB diffusion in bulk SiC does not properly describe the release observed from TRISO fuel.
		"Observations of Ag diffusion in ion implanted SiC" Tyler Gerczak, bin leng, Kumar Sridharan, Jerry Hunter, Andrew Giordani, Todd Allen, Journal of Nuclear Materials Vol. 461 2015 314-324
		"Observations of defect structure evolution in proton and Ni ion irradiated Ni-Cr binary alloys" Samuel A. Briggs, Khalid Hattar, Janne Pakarinen, Kumar Sridharan, Mitra Taheri, Christopher Barr, Mahmood Mamivand, Dane Morgan, Journal of Nuclear Materials Vol. Volume 479 2016 48-58 Link Two binary Ni-Cr model alloys with 5 wt% Cr and 18 wt% Cr were irradiated using 2 MeV protons at 400 and 500 °C and 20 MeV Ni4+ ions at 500 °C to investigate microstructural evolution as a function of composition, irradiation temperature, and irradiating ion species. Transmission electron microscopy (TEM) was applied to study irradiation-induced void and faulted Frank loops microstructures. Irradiations at 500 °C were shown to generate decreased densities of larger defects, likely due to increased barriers to defect nucleation as compared to 400 °C irradiations. Heavy ion irradiation resulted in a larger density of smaller voids when compared to proton irradiations, indicating in-cascade clustering of point defects. Cluster dynamics simulations were in good agreement with the experimental findings, suggesting that increases in Cr content lead to an increase in interstitial binding energy, leading to higher densities of smaller dislocation loops in the Ni-18Cr alloy as compared to the Ni-5Cr alloy.
		"Radiation-resistant nanotwinned austenitic stainless steel" Gabriel Meric, I.M. Robertson, Todd Allen, Jean-Claude van Duysen, Kumar Sridharan, Scripta Materialia Vol. 159 2019 123-127 Link A key strategy to increase the radiation resistance of materials has been to introduce a high density of interfaces that can act as sinks for radiation-induced defects. Twin boundaries are a type of interface that can be introduced through deformation but are usually considered to be ineffective sinks. Using heavy ion irradiation and transmission electron microscopy, this study investigates the influence of a high area per unit volume of twin boundaries on the radiation-induced swelling response of an austenitic stainless steel. The study shows that swelling can be suppressed in regions containing a high density of closely-spaced deformation twin boundaries.
		"Relationship between lath boundary structure and radiation induced segregation in a neutron irradiated 9wt.% Cr model ferritic/martensitic steel" Todd Allen, Heather Chichester, Kevin Field, Brandon Miller, Kumar Sridharan, Journal of Nuclear Materials Vol. 445 2013 143-148 Link Ferritic/Martensitic (F/M) steels with high Cr content posses the high temperature strength and low swelling rates required for advanced nuclear reactor designs. Radiation induced segregation (RIS) occurs in F/M steels due to solute atoms preferentially coupling to point defect fluxes which migrate to defect sinks, such as grain boundaries (GBs). The RIS response of F/M steels and austenitic steels has been shown to be dependent on the local structure of GBs where low energy structures have suppressed RIS responses. This relationship between local GB structure and RIS has been demonstrated primarily in ion-irradiated specimens. A 9 wt.% Cr model alloy steel was irradiated to 3 dpa using neutrons at the Advanced Test Reactor (ATR) to determine the effect of a neutron radiation environment on the RIS response at different GB structures. This investigation found the relationship between GB structure and RIS is also active for F/M steels irradiated using neutrons. The data generated from the neutron irradiation is also compared to RIS data generated using proton irradiations on the same heat of model alloy.
		"Size distribution of black spot defects and their contribution to swelling in irradiated SiC" Beata Tyburska-Puschel, Yizhang Zhai, Li He, Cheng Liu, Alexandre Boulle, Paul Voyles, Izabela Szlufarska, Kumar Sridharan, Journal of Nuclear Materials Vol. 476 2016 132-139 Link
		"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.
		"Structural Evolution of Oxidized Surface of Zirconium-Silicide under Ion Irradiation" Hwasung Yeom, Li He, Robert Mariani, Kumar Sridharan, Applied Surface Science Vol. 455 2018 333-342 Link
		"Study of Interfacial Interactions Using Thin Film Surface Modification" Alexander Mairov, Kumar Sridharan, Clarissa Yablinsky, Transactions of the American Nuclear Society Vol. 106 2012 1270 Link Interfaces play a key role in dictating the long-term stability of materials under the influence of radiation and high temperatures. For example, grain boundaries affect corrosion by way of providing kinetically favorable paths for elemental diffusion, but they can also act as sinks for defects and helium, generated during irradiation. Likewise, the stability of nanometer scale(Y, Ti)-oxide particles in nano-structured oxide dispersion strengthened(ODS) steels depends strongly on the stoichiometric and physical stability of the oxide particles/matrix interface under radiation and high temperatures[1]. A fundamental understanding of these interfacial effects is crucially important to the development of materials for extreme environments of a nuclear reactor. Research in the last three decades has shown that energetic ion beams can be powerful tools to modify near surface structure of materials[2]. For example, with heavy ions …

Presentations:

	"Cold Spray Coatings for Accident Tolerant Zr-Alloy Cladding in Light Water Reactors" Hwasung Yeom, Ben Maier, Peng Xu, Kumar Sridharan, ANS Annual Meeting 2018 June 18-22, (2018)
	"Complementary techniques for quantification of a' phase precipitation in neutron-irradiated Fe-Cr-Al model alloys" Philip Edmondson, Kevin Field, Kumar Sridharan, Microscopy & Microanalysis 2016 July 24-28, (2016) Link
	"Dependencies of a' Embrittlement in Neutron-Irradiated Model Fe-Cr-Al Alloys" Philip Edmondson, Kevin Field, Kumar Sridharan, 2016 ANS Annual Meeting June 12-16, (2016) Link
	"Effects of Ti and Al Additions on Irradiation Behavior of FeMnNiCr Based High Entropy Alloys" Andrew Hoffman, Haiming Wen, Li He, Kumar Sridharan, 2019 TMS Annual Meeting March 10-14, (2019)
	"Effects of Ti and Al Additions on Irradiation Behavior of FeMnNiCr Based High-Entropy Alloys" Matthew Luebbe, Andrew Hoffman, Hans Pommeranke, Li He, Kumar Sridharan, Haiming Wen, Materials Science & Technology 2019 September 29-3, (2019)
	"Investigation of High-Entropy Alloys Compositions for Radiation Damage Applications" Calvin Parkin, Michael Moorehead, Zefeng Yu, Kumar Sridharan, Adrien Couet, ANS Annual Meeting 2018 June 18-22, (2018)
	"Ion Irradiation Defects in Austenitic Alloy 709 and Ferritic-Martensitic Steel Grade 92 for Nuclear Applications" Li He, Rigen Mo, Beata Tyburska-Puschel, Kumar Sridharan, Haixuan Xu, Tianyi Chen, Lizhen Tan, MRS Spring 2017 April 17-21, (2017)
	"Ion irradiation for nuclear materials research at University of Wisconsin-Madison" Li He, Gabriel Meric, Kim Kriewaldt, Kumar Sridharan, Adrien Couet, Todd Allen, The 51st Symposium of the North Eastern Accelerator Personnel September 23-27, (2018)
	"Irradiation Effects in Oxide Nanoparticle Stability in Oxide Dispersion Strengthened (ODS) Steel" Todd Allen, Jianchao HE, Alexander Mairov, Kumar Sridharan, The Metallurgical Society Meeting March 15-19, (2015)
	"Manufacturing of nanostructured ODS steel cladding tubes for advanced nuclear reactors using cold spray technology" Mia Lenling, Hwasung Yeom, Ben Maier, Greg Johnson, David Hoelzer, Peter Hosemann, Stuart Maloy, Kumar Sridharan, Jeffrey Graham, The Minerals, Metals and Materials Society (TMS) 2019 March 10-14, (2019)
	"Microstructural Characterization of High-entropy Alloy Ion Irradiated at Cryogenic Temperatures" Michael Moorehead, Calvin Parkin, Lingfeng He, Jing Hu, Meimei Li, Adrien Couet, Kumar Sridharan, TMS 2019 March 10-14, (2019)
	"Program Review Presentation Entitled "Development of Low Temperature Spray Process for Manufacturing Fuel Cladding and Surface Modification of Reactor Components"" Mia Lenling, Kumar Sridharan, Hwasung Yeom, Peter Hosemann, David Hoelzer, Jeffrey Graham, Stuart Maloy, Advanced Methods for Manufacturing Program Review December 4-6, (2018)
	"Structural Effects in Oxide Dispersion Strengthened (ODS) Steel Neutron Irradiated to 3 dpa at 500°C" Jianchao HE, Alexander Mairov, Kumar Sridharan, M&M (Microscopy and Microanalysis) August 2-6, (2015)
	"Structural Effects in Oxide Dispersion Strengthened (ODS) Steel Neutron Irradiated to 3 dpa at 500°C" Jianchao HE, Alexander Mairov, Kumar Sridharan, Microscopy and Microanalysis 2015 meeting August 2-6, (2015)
	"Study of B2 and Laves Phase E volution in a Novel Ferr itic Steel under Ion Irradiation" Li He, Lizhen Tan, Ying Yang, Kumar Sridharan, MiNES (Materials in Nuclear Energy Systems) 2019 October 6-10, (2019)
	"Study of Interfacial Interactions Using Thin Film Surface Modification" Todd Allen, Alexander Mairov, Kumar Sridharan, MS&T Conference October 7-11, (2012)
	"Study of Interfacial Interactions Using Thin Film Surface Modification" Alexander Mairov, Kumar Sridharan, Clarissa Yablinsky, ANS Annual Meeting June 24-28, (2012)

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
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

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Reviewed 10+ RTE Proposals

NSUF Research Collaborations

Atom probe characterization of HT-9 as a function of neutron irradiation temperature - FY 2023 RTE 2nd Call, #4629

Atom Probe Tomography (APT) Investigation of Radiation Stability of Oxide Nanoclusters in Oxide Dispersion Strengthened (ODS) Steel Manufactured by the Cold Spray Process - FY 2019 RTE 2nd Call, #1731

Atomic Probe Tomography Studies of Irradiated Cold Spray Coatings for Accident Tolerant Cladding - FY 2017 RTE 3rd Call, #1069

Effect of neutron irradiation on the microstructure of NF616 (Grade 92) as a function of dose and temperature - FY 2022 RTE 1st Call, #4453

Enhanced irradiation tolerance of high-entropy alloys - FY 2017 RTE 3rd Call, #1122

Grain boundary microchemistry of ion irradiated Ferritic/Martensitic steels as determined by advanced microscopy techniques - FY 2011 RTE Solicitation, #313

in-situ Transmission Electron Microscopy during Ion-Irradiation Studies of Cold Spray Coatings for Accident Tolerant Cladding - FY 2018 RTE 3rd Call, #1519

Investigation of precipitate formation kinetics and interactions in FeCrAl alloys - FY 2015 RTE 2nd Call, #556

Ion irradiation of advanced materials – nanostructured steels and high entropy alloysNew Proposal - FY 2017 RTE 1st Call, #865

IVEM Investigation of Defect Evolution in Bulk High Entropy Alloys under Single- and Dual-beam Heavy-ion Irradiation - FY 2018 RTE 3rd Call, #1610

Mechanical characterization of neutron irradiated NF616 (T92) as a function of doses and temperatures - FY 2019 RTE 3rd Call, #2879

Mechanical characterization of three heats (ORNL, LANL and EBR II) of HT-9 after side-by-side neutron irradiation at LWR and fast reactor relevant temperatures - FY 2018 RTE 1st Call, #1156

Mechanical characterization of three lower dose HT-9 heats (ORNL, LANL and EBR II) after side-by-side neutron irradiation at LWR and fast reactor relevant temperatures - FY 2019 RTE 1st Call, #1687

Microstructural characterization of neutron irradiated HT-9 heats (ORNL, LANL and EBR II) at LWR and fast reactor relevant temperatures - FY 2020 RTE 2nd Call, #4201

Microstructural characterization of neutron irradiated NF616 (Grade 92) as a function of doses and temperatures - FY 2021 RTE 1st Call, #4259

Nanostructuring to enhance irradiation tolerance of ferritic/martensitic Grade 91 steels - FY 2018 RTE 2nd Call, #1403

Parametric study of factors affecting precipitation in model FeCrAl alloys - FY 2016 RTE 3rd Call, #687

Radiation Induced Segregation (RIS) in Ni-Cr Alloys - FY 2014 RTE 1st Call, #474

Sample Preparation for Ex-situ Transmission Electron Microscopy Study of Deformation-induced Twinning and Martensite in Two 316L Austenitic Stainless Steels: Role of Stacking Fault Energy and Grain Orientation - FY 2017 RTE 2nd Call, #933

The effectiveness of coherent and incoherent twin boundaries in alleviating radiation damage in heavy-ion-irradiated 316L austenitic stainless steels - FY 2017 RTE 3rd Call, #1105

Kumar Sridharan

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