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Cold Spray Materials Deposition
The cold spray research program consists of multiple projects for applications including accident tolerant fuels, production of free-standing cladding tubes, and repair of spent fuel canisters. Learn more about these projects below.
Alloys for Molten Salt Reactors and Salt Chemistry
Corrosion Protection of SiC-SiC Fuel Cladding
SiC-SiC composites are a leading candidate to replace Zr-alloy cladding in the next generation of LWRs. SiC has a wide range of favorable properties that allow for improved reactor performance and enhanced safety features compared to current Zr cladding. In this project we are investigating materials to protect the SiC-SiC cladding from gradual corrosion in the water coolant at normal operating conditions. Follow the link below to learn more about this project.
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High Temperature Tribology for Gas-Cooled Reactors
Alloys 800H and 617 are among the lead metallic materials being considered for structural applications in high temperature gas-cooled reactor (HTGR) which is expected to operate at temperatures at 750 °C or higher. Helium gas used as coolant in HTGR typically contains impurities such as H2, H2O, CO2, CO, and CH4 which can induce corrosion. Depending on the activity of carbon and the partial pressure of oxygen in helium, a variety of corrosion reactions can occur at the alloy surface including oxidation, carburization, and decarburization. Consequently, the tribological behavior of the alloys’ surfaces in rubbing components is expected to vary dramatically depending on the regime of environmental chemistry.
We have conducted research to characterize, understand, and model the elevated-temperature tribological behavior of alloys 800H and 617 in these various regimes of corrosion. Based on this understanding, strategically selected surface treatments such as aluminizing and shot peening were investigated for these alloys to mitigate their tribological damage in HTGR environments. The research was conducted in collaboration with Argonne National Laboratory.
 “High-Temperature Tribological Behavior of Structural Materials after Conditioning in Impure-Helium Environments for High-Temperature Gas-Cooled Reactor Applications” V. Pauly, C. Tesch, J. Kern, M. Clark, D. Grierson, D. Singh, O. Ajayi, and K. Sridharan, Journal of Nuclear Materials, 522, 2019, p. 311
 “Wear Performance of Incoloy 800HT and Inconel 617 in Various Surface Conditions for High-Temperature Gas-Cooled Reactor Components”,V. Pauly, J. Kern, M. Clark, D. Grierson, and K. Sridharan, Tribology International, 2020, 106715.
 “Effects of Aluminization via Thermo-Chemical Diffusion on Wear Behavior of Structural Materials for High Temperature Gas-Cooled Reactors”, J. Kern, V. Pauly, M. Clark, D. Grierson, and K. Sridharan, Metallurgical and Materials Transactions A, 52A, 2021, p. 2463.
Heat Transfer from Materials' Surfaces
In light water reactors, water boiling from materials’ surfaces determines efficiency and serves as a mechanism to cool the heated nuclear fuel cladding in accident conditions. Therefore, for safety and efficiency nuclear power plants good boiling performance on cladding surfaces is required. Heat transfer properties are measured by boiling heat transfer and critical heat flux.
A variety of surface modification technologies for zirconium-alloy fuel cladding has been studied in our group to enhance heat transfer performance and corrosion/oxidation resistance.The coatings include porous oxide/metallic nanoparticle coatings (Ti, TiO2, Al2O3, Y2O3) using nanofluid boiling and electrophoretic deposition, and dense metallic and ceramic coatings (MAX phase, Cr, FeCrAl, ZrSi2) using physical vapor deposition and cold spray deposition. The engineered materials’ surfaces altered surface characteristics such as wettability, roughness, capillary force, and oxygen affinity to improve overall heat removal performance from fuel cladding in normal operating and accident conditions. Collaborators on these projects included Idaho National Laboratory, Westinghouse Electric Company, Pohang University of Science and Technology in South Korea.
 H.Jo, H.Yeom, E.Gutierrez, K.Sridharan, M.Corradini, Evaluation of Critical Heat Flux of ATF Candidate Coating Materials in Pool Boiling, Nuclear Engineering and Design, Vol. 354, 110166, 2019.
Zirconium-silicide compounds have been considered as promising functional and structural materials in future nuclear reactor systems such as neutron reflectors in gas-cooled fast reactors (HTGR) and protective coating materials for Zr-alloy cladding in light water reactors (LWR). The materials are attractive as the neutron reflector in the light of their high melting points, excellent mechanical properties at high temperatures, and highly compatible elastic scattering cross-section with fast neutron spectrum. Furthermore, zirconium-silicides are robust in oxidative high temperature environments due to the formation of oxidation-resistant passive oxide layers of SiO2 and ZrSiO4 (zircon).
Oxidation of ZrSi2 sputter coatings on Zr-alloy substrates forms tenacious surface oxide films consisting of nanocrystalline/ amorphous ZrSiO4 and SiO2 at 700 °C and multilayered ZrO2 and SiO2 at 1000 °C, respectively. Moreover, past studies show that tenacious oxide layers of the zirconium- silicide formed at elevated temperatures provide engineering benefits as a structural material in oxidative and radiation environments. The structurally and compositionally homogenous microstructure resulting from radiation damage eliminated preferential corrosion sites where corrosion attack could potentially occur.
 H.Yeom, B.Maier, R.Mariani, D.Bai, K.Sridharan, Evolution of Multi-Layered Scale Structures during High Temperature Oxidation of ZrSi2, Journal of Materials Research, Vol. 31, n. 21, pp. 3409-3419, 2016
 H.Yeom, B.Maier, R.Mariani, D.Bai, S.Fronek, Peng Xu, K.Sridharan, Magnetron Sputter Deposition of Zirconium-Silicide Coating for Mitigating High Temperature Oxidation of Zirconium-Alloy, Surface and Coatings Technology, Vol. 316, pp. 30-38, 2017
 H.Yeom, C.Lockhart, R.Mariani, X.Bai, M.Corradini, K.Sridharan, Evaluation of Steam Corrosion and Water Quenching Behaviors of Zirconium-Silicide Coated LWR Fuel Claddings, Journal of Nuclear Materials, Vol. 499, pp. 256-267, 2018
 H.Yeom, L.He, K.Sridharan, Ion-irradiation response of complex oxide microstructure of zirconium-silicide, Applied Surface Science, Vol. 455, pp. 333-342, 2018
Fuel–Cladding Chemical Interactions (FCCI) in a nuclear reactor occur due to thermal and radiation enhanced interdiffusion between the cladding and fuel materials. This can have the detrimental effects of reducing the effective cladding wall thickness and the formation of low melting point eutectic compounds. Deposition of thin diffusion barrier coatings in the inner surface of the cladding can potentially reduce or delay the onset of FCCI.
The study examined the feasibility of using a nanofluid-based electrophoretic deposition (EPD) process to deposit coatings of Yttrium Stabilized Zirconia (YSZ), titanium, titanium oxide (TiO2), vanadium oxide (V2O3) as the diffusion barrier coating. The deposition parameters, including the nanofluid solvent, additive, particle size, current, and voltage were optimized using test flat substrates. Using a co-axial configuration in conjunction with the EPD process, the nanoparticles were successfully deposited uniformly on the inner surfaces of sections of cladding with 4 mm inner diameter. Such a coating is extremely hard to make by conventional coating technologies like thermal spray or vapor deposition. The diffusion couple experiment showed that the coatings significantly reduce the solid state interdiffusion between cerium (surrogate material of fuel pellet) to steel.
 Firouzdor, Vahid, et al. “Development of titanium diffusion barrier coatings for mitigation of fuel–cladding chemical interactions.” Surface and Coatings Technology, 219 (2013): 59-68.
 Firouzdor, Vahid, et al. “Development of yttrium stabilized zirconia (YSZ) diffusion barrier coatings for mitigation of fuel–cladding chemical interactions.” Journal of Nuclear Materials, 438.1-3 (2013): 268-277.
Emissivity of Reactor Materials
In the advanced high temperature Generation IV reactor concepts, heat transfer by radiation becomes an important consideration because of the fourth power dependence of radiated heat on temperature. The key material parameter that dictates the magnitude of heat radiated from the surface is emissivity. Emissivity is a surface phenomenon and is dictated by the materials’ surface chemical composition as well as the physical nature of the surface such as roughness and texture. Since corrosion of the surface will inevitably occur at high temperatures in the ambient coolant environments, it is important that studies on evaluation of emissivity for materials be closely integrated with the surface corrosion products that form on these materials’ surfaces.
We have investigated emissivity behavior of a wide range of reactor-relevant structural materials including ferritic steels, austenitic stainless steels, Ni-based alloys, and graphite. Trends in emissivity changes after exposure of alloys to molten salts, high temperature helium, liquid sodium and supercritical carbon-dioxide have been studied. First principles modeling of emissivity has been performed. Collaborators on these projects included Oak Ridge National Laboratory, Idaho National Laboratory, and University of Missouri.
 “Impact of Corrosion on the Emissivity of Advanced Reactor Structural Alloys”, J. L. King, H. Jo, A. Shahsafi, K. Blomstrand, K. Sridharan, and M. A. Kats, Journal of Nuclear Materials, 508, 2018, p. 465.
 “Spectral Emissivity of Oxidized and Roughened Metal Surfaces for Energy Systems”, H. Jo, J. L. King, K. Blomstrand, and K. Sridharan, International Journal of Heat and Mass Transfer, 115B, 2017, p. 1065.
 “Effects of Surface Roughness, Oxidation, and Temperature on the Emissivity of Reactor Pressure Vessel Alloys”, J. L. King, H. Jo, R. Tirawat, K. Blomstrand, and K. Sridharan, Nuclear Technology, 206, 1, 2017, p. 1.
 “Computation of Total Hemispherical Emissivity from Directional Spectral Models”, J. L. King, H. Jo, S. K. Loyalka, R. V. Tompson, and K. Sridharan, International Journal of Heat and Mass Transfer, 109, 2017, p. 894.
Radiation Damage-Tolerant Nanoceramic Coatings
This project was performed in collaboration with Italian Institute of Technology (IIT), Milan’s Nanoscience Center. In this study, amorphous alumina coatings (with nanocrystalline islands) were deposited using pulsed laser deposition (PLD) on stainless steel substrates to enhance corrosion resistance in liquid lead (-alloy) environment for the development of lead fast reactors (LFRs).
The coatings showed good corrosion resistance in this environment and surprisingly good ductility in U-bend tests. When ion irradiated to 180dpa, these amorphous coatings underwent a transformation to a nanocrystalline structure at above 20dpa which was accompanied by a decreasing hardness and an improvement in toughness. Detailed high-resolution transmission electron microscopy (HR-TEM) in the region just beneath nanohardness indents revealed the formation of amorphous bands suggesting this transformation to be a toughening mechanism. This research points to the feasibility of developing nanostructured materials that can potentially soften and toughen under radiation, rather than getting embrittled.
 “Helium Irradiation of Y2O3-Fe Bilayer System”, A. Mairov, D. Frazer, P. Hosemann, and K. Sridharan, Scripta Materialia, 162, 2019, p. 156.
 “Extreme Ion Irradiation of Oxide Nanoceramics: Influence of the Irradiation Spectrum”, F. García Ferré, A. Mairov, M. Vanazzi, Y. Serruys, F. Lepretre, L. Beck, L. Van Brutzel, A. Chartier, M.G. Beghi, K. Sridharan, and F. Di Fonzo, Acta Materialia, 143, 15, 2018, p. 156.
 “Radiation Tolerant Nanoceramic Coatings for Lead Fast Reactor Nuclear Fuel Cladding”, F.G. Ferré, A. Mairov, M.Vanazzi, S. Bassini, M. Utili, M. Tarantino, M. Bargaglia, F.R. Lamastra, F. Nanni, L. Ceseracciu, Y. Serruys, P. Trocellier, L. Beck, K. Sridharan, M.G. Beghi, and F. Di Fonzo, Corrosion Science, accepted (in press) 2017.
 “Radiation Endurance in Al2O3 Nanoceramics”, F. García Ferré, A. Mairov, L. Ceseracciu, Y. Serruys, P. Trocellier, C. Baumier, O. Kaïtasov, R. Brescia, D. Gastaldi, P. Vena, M.G. Beghi, L. Beck, K. Sridharan, and F. Di Fonzo, Nature Scientific Reports, DOI: 10.1038/srep33478 October, 2016.
Creep Crack Growth
Supercritical Carbon Dioxide Corrosion
The supercritical carbon-dioxide Brayton cycle (SC-CO2) is being considered for power conversion systems for a variety of advanced Generation IV nuclear reactor concepts which will operate at temperatures higher than the present light water reactors (LWR), as well as for high temperature fossil power and solar power plants. The SC-CO2 cycle provides higher efficiencies compared to the conventional steam cycle, and allows for smaller components size, and simpler cycle layout. However, long-term materials corrosion by oxidation and carburization in high temperature SC-CO2 is a concern.
Our group has investigated corrosion performance (including mechanisms of corrosion) of a wide range of alloys, including ferritic and austenitic steels, and nickel-alloys in SC-CO2 environments at temperatures up to 750 °C and 20 MPa (~3000 psi). We have also studied the role of trace impurities in CO2 on corrosion, potential corrosion effects in solid state diffusion bonds required for manufacture of compact heat exchangers, mechanical stability of oxide layers, and have modeled long-term corrosion behavior of alloys in SC-CO2 environments. Collaborators on this project included Sandia National Laboratories, and Atomic Energy Commission or CEA, France.
 “Effect of Supercritical CO2 on the Performance of 740H Fusion Welds”, A. Brittan, J. Mahaffey, M. Anderson, and K. Sridharan, Materials Science & Engineering A, 742, 2019, p. 414.
 “Effect of CO and O2 Impurities on Supercritical CO2 Corrosion of Inconel Alloy 625”, J. Mahaffey, A. Schroeder, D. Adam, A. Brittan, M. Anderson, A. Couet, and K. Sridharan, Metallurgical and Materials Transactions A, 49A, 2018, p. 3703.
Supercritical Water Corrosion
At temperatures and pressures above 373 °C and 22 MPa, normal water undergoes a phase change and becomes supercritical (often referred to as the fourth state). Supercritical water (SCW) has superior heat transfer properties but can be quite corrosive to materials. Supercritical water-cooled reactor (SCWR) was among the Generation IV reactor concepts. Between years 2000 and 2008 our research focused extensively on advancing the fundamental understanding of mechanisms of corrosion, effects of dissolved oxygen on corrosion, and oxide spallation effects in water environments at temperatures and pressures up to 650 °C and 25 MPa. The research established corrosion criteria for broad classes of candidate ASME codified alloys, including ferritic steels, austenitic stainless steels, and Ni-based superalloys, and resulted in a comprehensive performance ranking for the use of these alloys in SCW environments.
Among the innovative achievements of this research were surface-modification and coating treatments of ferritic steels to enhance corrosion resistance by fundamentally changing the composition and structure of the surface oxide layer, grain boundary engineering to mitigate spallation of the protective oxide layer in austenitic steels, and shot peening to increase oxidation resistance by promoting chromium diffusion to the surface through grain refinement. The corrosion research conducted by our group contributed to many SCWR corrosion programs world-wide, including those in China, Canada, Japan, and South Korea. Collaborators for this project included Idaho National Laboratory, Argonne National Laboratory, University of Michigan, and Korea Atomic Energy Research Institute (KAERI).
 “Oxidation Behavior of Grain Boundary Engineered Alloy 690 in Supercritical Water Environment”, P. Xu, L. Y. Liang, K. Sridharan, and T. R. Allen, Journal of Nuclear Materials, 422, 1, 2012, p. 143.
 “Corrosion of Alumina-forming Austenitic steel Fe-20Ni-14Cr-3Al-0.6Nb-0.1Ti in Supercritical Water”; S.Nie, Y. Chen, X. Ren, K. Sridharan, and T.R. Allen, Journal of Nuclear Materials,399, 2010, p.231.
 “Effect of Grain Refinement on Corrosion of Ferritic-Martensitic Steels in Supercritical Water Environment”; X. Ren, K. Sridharan, and T.R. Allen, Materials and Corrosion-Werkstoffe und Korrosion, vol. 61, 9, 2010, p.748.
 “Effect of Shot Peening on the Oxidation of Alloy 800H Exposed to Supercritical Water and Cyclic Oxidation”; L. Tan, X. Ren, K. Sridharan, and T.R. Allen, Corrosion Science, vol. 50, 2008, p. 2040.
 “Corrosion Behavior of Ni-Based Alloys for Advanced High Temperature Water-Cooled Nuclear Plants”; L. Tan, X. Ren, K. Sridharan, and T.R. Allen, Corrosion Science, vol. 50, 2008, p. 3056.
 “The Role of Grain Boundary Engineering in the SCC behavior of F-M alloy HT-9 in Supercritical Water”; G. Gupta, P. Ampornrat, X. Ren, K. Sridharan, T. R. Allen, and G. S. Was, Journal of Nuclear Materials, vol. 361, 2007, p. 160.
 “Corrosion Behavior of Alloys 625 and 718 in Supercritical Water”; X. Ren, K. Sridharan, and T.R. Allen, Corrosion, vol. 63, 7, 2007, p. 603.