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Sridharan, Kumar

Publications and source records attributed to Sridharan, Kumar.

At least 73 records · Page 4
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DOE OSTI2019

Evolution of B2 and laves phases in a ferritic steel under Fe 2+ ion irradiation at 475 °C

Microstructural evolution in a novel ferritic steel (Fe–12Cr–3W–3Ni–3Al–1Nb, in wt.%) computationally designed to contain B2 and Laves phases after 4 MeV Fe 2+ ion irradiation up to 220 dpa at 475 °C was characterized using transmission electron microscopy in conjunction with x-ray energy dispersive spectroscopy. The ferritic matrix phase exhibited dislocation loops and tangled dislocations, but our focus was on stability of two types of intermetallic precipitates. The B2–NiAl precipitates ~13 nm in size remained crystalline and appeared to have slightly lower Al concentration after irradiation. The Laves phase, (Fe,Cr) 2 (Nb,W), were present in two size ranges: coarse micron-scale precipitates which were amorphized with a slight composition change at irradiation doses above~30 dpa, while the finer precipitate particles~100 nm in size were partially disintegrated with a noticeable composition change at doses above~70 dpa. Meanwhile, many Nb/Cr-enriched particles ~8 nm in size formed within a few hundreds of nanometers from the disintegrated particles. The understanding of the phase stability would help design advanced steels and engineer microstructures that are stable against high irradiation doses, while retaining good high temperature strength.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Evaluation of critical heat flux of ATF candidate coating materials in pool boiling

The critical heat flux (CHF) of potential accident tolerant fuel (ATF) cladding materials and concepts, Cr and FeCrAl coatings on zirconium-alloy substrates, was investigated experimentally. A cold spray process, which involves the propulsion of high velocity powder particles on to the substrate surface to form a coating upon impact, was used to create the coatings. Surface parameters such as contact angle, surface roughness, and morphology that are associated with CHF in pool boiling were characterized. Non-hydrodynamic effects of newly developed ATF candidate coating materials on CHF were also evaluated. The effects of thermal properties and coating thickness and expected challenges for evaluation of CHF for new ATF cladding materials development are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Manufacturing Oxide Dispersion-Strengthened (ODS) Steel Fuel Cladding Tubes Using the Cold Spray Process

The cold spray materials deposition process has been explored for manufacturing oxide dispersion-strengthened (ODS) steel fuel cladding tubes. Gas-atomized 14YWT ODS steel powder was used as the feedstock material. A parametric investigation of the cold spray process involving substrate materials of various hardnesses, gas preheat temperatures, and carrier gas compositions was performed to gain the highest quality deposit. The high-velocity impact of the powder on the substrate led to dissolution of discrete oxide nanoparticles, which subsequently reprecipitated during postdeposition annealing at high temperatures. The tubes were manufactured by deposition on an Al-alloy mandrel substrate and subsequent chemical dissolution of the substrate. A 204-mm-long and 1-mm-thick ODS steel cladding tube was successfully manufactured. The grain growth and distribution of oxide nanoparticles in ferritic steel matrix were identified at elevated temperatures. In general, the cold spray process holds considerable promise for rapid, cost-effective manufacturing of ODS steel cladding tubes.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Radiation response of Ti 2 AlC MAX phase coated Zircaloy-4 for accident tolerant fuel cladding

A Ti 2 AlC MAX phase material was deposited on Zircaloy-4 substrate using the cold spray process, with the goal of enhancing the accident tolerance of zirconium-alloy fuel cladding in light water reactors. The samples were annealed up to 923 K to study the intermetallic phase formation between the coating and Zircaloy-4 substrate. Three interface compounds, ZrAl 2 , ZrAl, and Zr 3 Al, were identified and the activation energy of the interfacial intermetallic compound growth was determined. The radiation response of the coating material and interfacial intermetallic compounds after 3.5 MeV Zr 2+ ion irradiation was investigated by transmission electron microscopy and nanoindentation. Ion irradiation disrupted the nanolamellar structure of the Ti 2 AlC coating while amorphization was observed some of the intermetallic compounds. Furthermore, nanoindentation revealed overall hardening of the Ti 2 AlC coating and Zr 3 Al phase, with no significant change observed in the ZrAl 2 and ZrAl phases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Interface reactions and mechanical properties of FeCrAl-coated Zircaloy-4

FeCrAl has been proposed as a coating layer to enhance accident tolerance of Zircaloy-4 fuel cladding. In this study, we systematically examined the diffusion kinetics and interface compound formation of the spray-coated FeCrAl and Zircaloy-4 substrate. After vacuum annealing at 923 K, 963 K, and 998 K for various time periods, a mixed Zr 2 Fe phase and Zr 3 Fe formed at the interface. The formation energies of these two phases were found to be 65 kJ/mol and 58 kJ/mol for Zr 2 Fe and Zr 3 Fe, respectively. Bending tests of cantilevers prepared by a focused ion beam technique showed good interface bonding between the coating layer and the substrate. Nanoindentation and micropillar compression tests in the cross-section of polished samples show different mechanical responses of each phase. The substrate exhibited the lowest hardness and most ductile deformation under compression, while interfacial Zr 2 Fe and Zr 3 Fe possessed the highest hardness and brittle deformation.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Report on The Down-Selected Nanoprecipitates-Strengthened Advanced Ferritic Alloys for Nuclear Reactor Applications

One ferritic steel (BL-Nb) and one ferritic-martensitic steel (TTZ2M) were down-selected for further investigation. Based on the TTZ2M, two large-scale ferritic-martensitic heats, i.e., FTT1 and FTN1, were produced at industrial steelmaking facilities, which are in contrast to the small laboratory heats BL-Nb and TTZ2M. Other than the standard normalization and tempering condition of the ferritic-martensitic steels, a heat-treated condition was explored for the FTT1 and FTN1, which is named as FTT1-T and FTN1-T, respectively. Two-step aging was applied to BL-Nb to favor the precipitation of Laves and B2-NiAl phases. The alloy compositions and heat treatment temperatures were guided by computational thermodynamics and dilatometry. Tensile, Charpy impact, and creep tests were conducted on the samples of the steels in air, using a commercial heat of T91 as a reference. Microstructural characterization was primarily using transmission electron microscopy and energy dispersive x-ray spectroscopy with occasional uses of optical microscopy.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Development of nanoprecipitates-strengthened advanced ferritic alloys for nuclear reactor applications

By using the experimentally justified and optimized thermodynamic database, five ferritic steels and three ferritic-martensitic steels were designed for property screening, including Vickers hardness, tensile, creep, and radiation hardening, which led to the down-selection of one ferritic steel (BL-Nb) and one ferritic-martensitic steel (TTZ2M) for further investigation. Based on the TTZ2M, two large-scale ferritic-martensitic heats, i.e., FTT1 and FTN1, were produced at an industrial steelmaking facility, which are in contrast to the small laboratory heats BL-Nb and TTZ2M. Other than the standard normalization and tempering condition of the ferritic-martensitic steels, a heat-treated condition was explored for the FTT1 and FTN1, which is named as FTT1-T and FTN1-T, respectively. Two-step aging was applied to BL-Nb to favor the precipitation of Laves and B2-NiAl phases. Tensile, Charpy impact, and creep tests were conducted on the samples of the steels in air, using a commercial heat of T91 as a reference. Microstructural characterization was primarily conducted using transmission electron microscopy and energy dispersive x-ray spectroscopy with occasional uses of optical microscopy and scanning electron microscopy.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Application of voltammetry for quantitative analysis of chromium in molten 2LiF-BeF 2 (FLiBe) salt

The electrochemical behavior of Cr 2+ in molten 2LiF-BeF 2 (FLiBe) salt at temperatures between 500 °C and 700 °C was studied using cyclic voltammetry (CV) and linear sweep voltammetry (LSV). LSV has been used to establish a stable reference electrode with the Be/Be 2+ couple. The ability to distinguish between common corrosion products, such as Cr2+ and Fe 2+ , with CV has been demonstrated using the Be/Be 2+ reference reaction. Methods for high-precision, in-situ quantitative measurement of Cr 2+ concentration are developed. Models for predicting the concentration of Cr 2+ and its diffusion coefficient are approximated by relating faradaic cathodic peak current density to Cr 2+ concentration at 600 °C and 650 °C. Here this study provides the basis for the development of a sensor for on-line quantification of corrosion products that improves understanding of materials performance in the molten FLiBe salt.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Natural circulation FLiBe loop overview

A new large natural circulation FLiBe loop has operated for the first time in 40 years. Reaching temperatures above 700 C with stable flow rates around 5 cm/s, the loop exhibits a range of flow phenomena pertinent to high temperature natural circulation safety systems for nuclear reactors. This paper discusses the system and several operating conditions spanning a wide range of behavior to aid the development of other systems, simulation benchmarks, or system-codes. Analysis of heat transport phenomena indicates that standard Nusselt number correlations apply to a first approximation. More accurate salt-freezing models may be necessary for air-cooled salt-systems due to the insulating affect of the low thermal-conductivity frozen salt layer on the cold pipe walls.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Integrated Computational Study of Radiation Damage Effects in Grade 92 Steel and Alloy 709

Radiation resistance of austenitic alloy 709 (named as A709) and ferritic-martensitic steel Grade 92 (G92) were investigated in this project using 316H and Grade 91 (G91) as references, respectively. In the meantime, a model 709 (M709) alloy (Fe-20Cr-24Ni) was prepared to study the matrix response to irradiation. Samples machined from the tab section of three creep-ruptured A709 samples, i.e., 17,106 h at 600 °C, 12,014 h at 650 °C, and 16,448 h at 700 °C, were included in this work to evaluate radiation effect on the long-term “thermally aged” samples. Primarily Fe 2+ ion irradiation was conducted at 350, 600 and 670 °C for up to 75 displacements per atom (dpa), together with a few proton irradiation experiments at 380–670 °C for up to 1.5 dpa. Additionally, four neutron-irradiated NF709 samples (3.22–8 dpa at ~430–469 °C) were examined. The samples were examined by transmission electron microscopy (TEM), scanning electron microscopy (SEM) and energy dispersive x-ray spectroscopy (EDS) for microstructures and nanoindentation for radiation-hardening. Unlike the negligible hardening after irradiation at 670 °C, irradiation to nominal 2.5–50 dpa at 350 °C resulted in radiation-hardening of 13.5±5.4 % for A709, which was comparable to that of 13.4±6.2 % for M709 and moderately lower than that of 20.1±5.6 % for 316H. The radiation-hardening induced by the same irradiation condition to the “thermally aged” samples was comparable to A709, which were hardened by 15.0±8.2 %, 13.5±6.0 %, and 10.1±6.1 % for the 600, 650, and 700 °C aged samples, respectively. The radiation-hardening of A709 is significantly lower than that (~94±9 %) of neutron-irradiated NF709 to 8 dpa at ~430 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Integrated Computational Study of Radiation Damage Effects in Grade 92 Steel and Alloy 709

Radiation resistance of austenitic alloy 709 (named as A709) and ferritic-martensitic steel Grade 92 (G92) were investigated in this project using 316H and Grade 91 (G91) as references, respectively. In the meantime, a model 709 (M709) alloy (Fe-20Cr-24Ni) was prepared to study the matrix response to irradiation. Samples machined from the tab section of three creep-ruptured A709 samples, i.e., 17,106 h at 600 °C, 12,014 h at 650 °C, and 16,448 h at 700 °C, were included in this work to evaluate radiation effect on the long-term “thermally aged” samples. Primarily Fe 2+ ion irradiation was conducted at 350, 600 and 670 °C for up to 75 displacements per atom (dpa), together with a few proton irradiation experiments at 380–670 °C for up to 1.5 dpa. Additionally, four neutron-irradiated NF709 samples (3.22–8 dpa at ~430–469 °C) were examined. The samples were examined by transmission electron microscopy (TEM), scanning electron microscopy (SEM) and energy dispersive x-ray spectroscopy (EDS) for microstructures and nanoindentation for radiation-hardening

36 MATERIALS SCIENCE↗
DOE OSTI2018

In situ TEM investigation of irradiation-induced defect formation in cold spray Cr coatings for accident tolerant fuel applications

The first in-situ ion irradiation and transmission electron microscopy (TEM) of cold spray deposited Cr coatings was performed at the Intermediate Voltage Electron Microscope (IVEM) facility at Argonne National Laboratory for preliminary investigation into the radiation damage tolerance of these materials for use in accident tolerance fuel cladding. The severely plastically deformed microstructure of the cold spray deposited Cr delayed the onset and growth of radiation induced defects when compared to an annealed coating simulating bulk Cr. Evidence to support this conclusion was based on defect density and defect size measurements of TEM images obtained at different irradiation intervals.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗