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Sridharan, Kumar

Publications and source records attributed to Sridharan, Kumar.

At least 55 records · Page 3

Experimental evaluation of direct current magnetron sputtered and high-power impulse magnetron sputtered Cr coatings on SiC for light water reactor applications

Two types of physical vapor deposition Cr coatings were deposited on chemical vapor deposited SiC coupons using conventional direct current magnetron sputtering (DCMS) and high-power impulse magnetron sputtering (HiPIMS) to mitigate the corrosion of SiC, an advanced fuel and structural material in light water nuclear reactor cores. Coating microstructure was characterized using scanning electron microscopy and scanning transmission election microscopy. X-ray diffraction analysis was conducted for phase identification and qualitative analysis of residual stresses. The mechanical integrity of the coatings was evaluated using scratch testing, micro-indentation, and nano-indentation tests. Additionally, the microstructure of the DCMS coating consisted of fine columnar grains with nanoscale intercolumnar channels, while the HiPIMS microstructure was denser and free of any columnar defects. The DCMS coating showed a small tensile residual stress, while the HiPIMS coating was under compressive stress. Both types of coatings showed good mechanical integrity in terms of ductility and adhesion to the substrate. Corrosion performance of the coatings was assessed with a 30-day high temperature water autoclave test. Both types of coatings formed a protective Cr 2 O 3 surface layer 20-30 nm thick during the autoclave test. In the DCMS coating, the columnar defects provided permeation pathways for water penetration resulting in oxide formation in the intercolumnar regions in the interior of the coating. Overall, both types of Cr coatings offer promise for the mitigation of hydrothermal corrosion of SiC in light water reactor operating environments.

36 MATERIALS SCIENCE↗

Pool boiling critical heat flux studies of accident tolerant fuel cladding materials

Pool boiling experiments have been performed at atmospheric pressure on potential accident tolerant fuel (ATF) cladding materials (Zirlo®, Cr coated Zirlo®, FeCrAl coated Zirlo®, and monolithic SiC) to evaluate the Critical Heat Flux (CHF) points. Corrosion resistant coatings of Cr and FeCrAl alloy were deposited on flat Zirlo® samples using cold spray technology. The as-prepared samples after uniform surface polishing were subjected to autoclave tests at 360 °C water and 18.6 MPa for 360 h to simulate prototypic corrosion of the ATF cladding materials in LWR normal operation. Surface characteristics potentially influencing CHF such as surface morphology, roughness, static contact angle, and surface chemistry were characterized using a suite of characterization methods including scanning electron microscopy, 3D optical profilometry, optical contact angle measurements, and x-ray photoelectron spectroscopy (XPS). Thermo-physical properties of the samples such as density, thermal conductivity, and heat capacity were measured by differential scanning calorimetry and laser flash thermal diffusivity measurement. FeCrAl coatings showed CHF values comparable to bare Zirlo® samples, while slightly lower CHF values were observed for Cr coated samples and bulk SiC flats (produced by chemical vapor deposition, CVD). As-deposited Cr coatings showed 67% higher CHF than the as-polished Cr coatings due to its higher surface roughness level. The autoclave testing for the prepared samples generally increased the CHF except for the surface polished Cr coatings. Similar CHF values of bare Zirlo®, FeCrAl coating, and CVD SiC samples were observed after autoclave test but there were negligible change CHF of Cr coatings. XPS studies indicated that the formation or deposition of a few monolayers of hydrophobic carbonaceous species on surface can affect CHF values. The trends in CHF data are discussed in terms of evolution of the surface characteristics of the ATF materials.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Method for identification of redox control parameters for corrosion mitigation in molten fluoride salts

Structural material corrosion is a significant concern for the development of molten salt-cooled reactors (MSR) and redox control was proposed to mitigate the corrosion. This study presents an efficient approach to identify the partial pressure of F 2 gas for fluorine driven corrosion, the concentration ratios of UF 4 /UF 3 and ZrF 4 /ZrF 2 , and the partial pressure ratio of HF/H 2 needed to limit metal corrosion and its subsequent dissolution in molten fluoride salts. The identification procedure is made by simply using auxiliary axes in Ellingham diagrams. This approach greatly simplifies the optimization of redox control methods in studying the material corrosion in molten salts.

36 MATERIALS SCIENCE↗

Wear Performance of incoloy 800HT and inconel 617 in various surface conditions for high-temperature gas-cooled reactor components

Incoloy™ 800HT and Inconel™ 617 are primary candidate structural materials for high-temperature gas-cooled reactors (HTGRs). Impurities (e.g., H 2 O, CH 4 ) in the coolant induce corrosion reactions that affect tribological behavior. We investigate the high-temperature tribological performance of the alloys under surface conditions including: conditioned in H 2 O-containing helium at elevated temperatures, hardened via the carbon-diffusion Kolsterising® treatment, and conditioned after hardening. Initial friction coefficients and volume removed serve as metrics for wear performance. In this work, we find that wear mechanisms including scuffing, glaze-oxide formation, and oxide-layer wear and break-through are governed by surface hardness, chemistry, and applied load. The formation of a glaze-oxide layer is achieved with an oxide of sufficient mechanical integrity if the oxide-substrate interface is strong and the load is low.

42 ENGINEERING↗

Cold spray technology in nuclear energy applications: A review of recent advances

Cold spray technology is a promising solid-state, powder-based deposition methods for fabrication of coatings, near-net-shape manufacturing, and component repair. In recent years, this technology has been investigated for many applications in the nuclear energy sector. For example, it has been explored for the deposition of coatings of corrosion and oxidation materials on zirconium-alloys fuel claddings in light water reactors (LWR) for achieving more accident-tolerant fuel (ATF) cladding. Other examples of its use for nuclear energy applications include coatings for enhancements in wear resistance and heat transfer, near-net shape manufacturing of next-generation fuel claddings of oxide dispersion-strengthened (ODS) steels, and for mitigation and repair of potential chloride induced stress corrosion cracking in used fuel storage canister systems. This work reviews published literature on the application of cold spray process for nuclear energy applications, while pointing to gaps that have yet to be addressed.

36 MATERIALS SCIENCE↗

Cold Spray Deposition of 304L Stainless Steel to Mitigate Chloride-Induced Stress Corrosion Cracking in Canisters for Used Nuclear Fuel Storage

A feasibility study of the cold spray deposition process of 304L stainless steel on 304L stainless steel substrates as a mitigation method for chloride-induced stress corrosion cracking (CISCC) has been investigated under various substrate conditions. The study is aimed at the application of this technology to mitigate CISCC that may potentially occur in or nearthe fusion welded regions of stainless steel canisters in Dry Cask Storage System (DCSS) for used nuclear fuels. Spherical gas-atomized 304L stainless steel powder, in the size range of 25 µm to 44 µm, was used as a feedstock powder for the cold spray process. The powder was deposited on four types of 304L substrates with various surface conditions: as-polished, oxidized, cold-rolled, plates with prototypical CISCC. The effects of cold spray parameters on quality of cold spray coatings were investigated. Thickness, porosity, and phases in the as-deposited materials were evaluated using scanning electron microscopy (SEM) and X-ray diffraction (XRD), and correlated with microhardness and adhesion strength measured via micro-indentation and ASTM C633 pulling test, respectively. XRD analysis of the coatings was also conducted to examine the effects of cold spray condition on residual stress state in the coating. Detailed cross-sectional examination of coating/substrate interfaces was performed with transmission electron microscopy (TEM) equipped with energy dispersive spectroscopy (EDS). Dense and continuous coatings with good adhesion strength and hardness were produced for the various substrate conditions by adjusting cold spray parameters. The results demonstrate that cold spray stainless steel coating is a viable option to provide a physical barrier against CISCC in fusion weld regions of stainless steel in corrosive chloride salt-bearing environments.

Dry Cask Storage Systems, cold spray, 304L stainle↗

Non-galvanic mass transport in molten fluoride salt isothermal corrosion cells

Mass transport of corrosion products between Ni-201 test samples and 316 L capsules in a 500-hour FLiNaK corrosion experiment has been investigated. In one experiment, the capsule and samples were electrically connected for galvanic coupling, while in the other they were electrically isolated using boron-nitride. Post-corrosion characterization of Ni-201 from the two experiments did not reveal any remarkable differences. Substantial transport of Fe and Cr from the 316 L to the Ni-201 via the salt was observed, with notable diffusion of these elements in the near-surface regions of the Ni-201 samples, while a Cr-rich film was also observed on their surfaces.

36 MATERIALS SCIENCE↗

In situ microstructural evolution in face-centered and body-centered cubic complex concentrated solid-solution alloys under heavy ion irradiation

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 mechanisms 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 various levels of displacements per atom (dpa) using in-situ transmission electron microscopy. Cryogenic irradiation 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 kinetics 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.

36 MATERIALS SCIENCE↗

Radiation response of FeCrAl-coated Zircaloy-4

Coating the surface of Zircaloy-4 light water reactor fuel cladding tube is of considerable interest for enhancing its accident tolerance. In this study, thermal annealing of FeCrAl-coated Zircaloy-4 at 725 °C for 500 h, was used to induce interfacial reactions between Zircaloy-4 and FeCrAl. The interface zones were then irradiated by 3.5 MeV Zr ions at 400 °C, up to 50, 100, 150 peak dpa values. Transmission electron microscopy (TEM) and scanning TEM were used to characterize microstructural and composition changes before and after ion irradiation. Three interfacial phases were identified: FeZr 3 , (Fe,Cr) 2 Zr, and ZrC. The widest intermetallic layer, FeZr3, had large grains. The narrower phases, (Fe,Cr)2Zr and ZrC, contained small grains and were often mixed. The unexpected observation of the ZrC phase was attributed to the presence of very small impurity level concentrations of carbon in the powder material. No void swelling was observed in any of the phases, including the FeCrAl coating and Zircaloy-4 substrate. (Fe,Cr) 2 Zr, however, fully amorphized after irradiation, even at the lowest dpa.

36 MATERIALS SCIENCE↗

Investigation of impurity driven corrosion behavior in molten 2LiF-BeF 2 salt

A 1000 -h corrosion experiment of 316H stainless steel at 700 °C investigated the role of metal fluoride impurities CrF 2 and FeF 2 in LiF-BeF 2 (66-34 mole %) (FLiBe) salt. Inductively coupled plasma-optical emission spectroscopy (ICP-OES) demonstrated the ability to measure changes in the concentration of these fluorides in FLiBe salt. Cyclic voltammetry (CV) showed the potential for real-time quantitative concentration measurement of corrosion products in the molten salt. Materials characterization revealed that both impurity species influence the dissolution of Cr from the alloy. The collective results from these analysis methods advance the understanding of the influence impurities have on materials corrosion.

2LiF-BeF2↗

Tribological Behavior of Structural Materials in High Temperature Helium Gas-Cooled Reactor Environments

The High-Temperature Gas-cooled Reactor (HTGR) is a Generation IV concept designed to produce electricity and hydrogen at high efficiencies via high operating temperatures (700 °C or higher). Helium, the primary coolant for HTGRs, contains impurities (e.g., CO, H2O and CH4) that can induce corrosion reactions at high temperatures such as oxidation and (de)carburization, which in turn can affect the tribological behavior of components. Incoloy® 800HT (Ni-Fe-Cr austenitic solid-solution alloy) and Inconel® 617 (Ni-Cr-Co-Mo solid-solution alloy) are two high-temperature superalloys currently selected as candidate structural materials for the HTGR. The objective of this study is to evaluate the high-temperature tribological performance of these two candidate alloys after conditioning them in HTGR environments. Four regimes of corrosion were considered: a non-conditioned regime (Regime I), an oxidizing regime (Regime II), a carburizing and oxidizing regime (Regime III), and a carburizing regime (Regime IV). To simulate oxidation in HTGR, samples were conditioned for 22 days at elevated temperatures in a once-through helium loop with 4 ppmv H2O. Carburization of the samples was achieved by the commercial process, Kolsterising®. In addition, two surface treatments – shot peening and aluminization – were considered as potential routes of improving the alloys’ wear resistance. Surface-treated samples were tested both before and after conditioning in Regime II. Tribological testing of the samples was performed via a pin-on-disk tribometer at elevated temperatures in ambient environment – 650 and 750°C for 800HT; 850 and 900°C for 617 – with applied loads of 1N, 2N and 5N. The wear behavior of the alloys was assessed via wear volume and friction coefficient measurements, supported by morphological, structural and compositional analyses of the wear tracks. Conditioning the samples in Regime II led to the formation of a chromium oxide on both alloys. This protective scale increased the wear resistance compared to that of as-received samples (Regime I) due to the formation of a compacted ‘glaze’ oxide layer during sliding, rendering the wear track nearly undistinguishable from the unworn background and resulting in wear volumes below the detection limits of the measurement technique. The variability of the friction coefficients of the conditioned samples was also considerably reduced compared to that of the as-received samples due to the glaze layer. Additionally, the initial friction coefficients of the samples conditioned in Regime II were reduced by 45% and 54% compared to those of as-received samples for alloys 800HT and 617, respectively. Carburizing the samples (Regime IV) hardened the surfaces of both alloys and promoted the formation of an iron oxide on 800HT during tribological testing, thereby increasing the wear resistance and decreasing the initial friction coefficient by a factor of two compared to those of as-received 800HT. Alloy 617 exhibited an enhanced wear resistance but similar initial friction coefficient compared to those of as-received samples due to the increased surface hardness. The Mn-Cr oxide developed during Regime III conditioning of 800HT did not develop a glaze layer during tribotesting, even at lower loads. Thus, it was not as protective as the oxide produced during conditioning of Regime II samples, leading to lower wear resistance of 800HT in Regime III compared to that measured in the Regime II condition. 800HT benefited from the aluminization surface treatment, particularly before conditioning in Regime II, due to the promotion of a wear-resistant aluminum oxide layer during testing and the increase in surface hardness due to the presence of a FeAl intermetallic phase. The presence of this phase caused an order-of-magnitude reduction in wear volumes compared to those of as-received 800HT. Aluminization of 617 decreased its wear resistance, particularly after conditioning in Regime II, due to significant wear of the aluminum oxide via a ceramic wear mechanism. This aluminum oxide was removed during sliding, causing the wear volumes to increase by a factor of five compared to those of 617 conditioned in Regime II. Shot-peened 800HT exhibited a tribological behavior similar to that of as-received 800HT, despite an enhancement in the rate of chromium oxide formation. Shot-peened 800HT conditioned in Regime II showed lower wear resistance compared to that of 800HT conditioned in Regime II due to the poor adhesion between the oxide and the underlying metal, preventing the formation of the glaze layer during sliding. Shot peening of 617 increased its wear resistance at higher load, reducing worn volumes by 60% compared to that measured with as-received 617. Tribotesting of shot-peened 617 in Regime II revealed that the glaze-oxide layer was formed at every load, resulting in negligible wear, similar to what was observed from testing 617 after conditioning in Regime II. Overall, alloy 617 exhibited tribologically superior behavior compared to that of alloy 800HT, as demonstrated by its lower friction coefficients and its order-of-magnitude lower wear volumes when measurable.

36 MATERIALS SCIENCE↗

Cold spray chromium coating for nuclear fuel rods

A method is provided for coating the substrate of a component, such as a zirconium alloy cladding tube, for use in a water cooled nuclear reactor under normal operating conditions and under high temperature oxidation conditions. The method includes heating a pressurized carrier gas to a temperature between 200° C. and 1200° C., adding chromium or chromium-based alloy particles having an average diameter of 20 microns or less to the heated carrier gas, and spraying the carrier gas and particles onto the substrate at a velocity, preferably from 800 to 4000 ft./sec. (about 243.84 to 1219.20 meters/sec.), to form a chromium and/or chromium-based alloy coating on the substrate to a desired thickness.

36 MATERIALS SCIENCE↗

Corrosion and Thermal Stability of CrMnFeNi High Entropy Alloy in Molten FLiBe Salt

The corrosion behavior of the FCC Cr 18 Mn 27 Fe 27.5 Ni 27.5 high entropy alloy (HEA) after exposure to molten FLiBe salt at 700 °C for 1000 hours, has been investigated. Results show that the HEA lost a higher mass compared to the reference 316 H stainless steel due to the dissolution of Mn into the molten salt. The loss of Mn from the alloy appeared to discourage the dissolution of Cr in the molten fluoride salts which is widely recognized as the mechanism of corrosion degradation. Thermal exposure at 700 °C for 1000 hours also led to the precipitation of an additional BCC phase Cr 67 Fe 13 Mn 18.5 Ni 1.5 , which was confirmed by CALPHAD predictions.

36 MATERIALS SCIENCE↗

High-temperature Tribological Behavior of Structural Materials after Conditioning in Impure-Helium Environments for High_temperature Gas-Cooled Reactor Applications

Incoloy 800HT and Inconel 617 have been selected as candidate structural alloys for the hightemperature gas-cooled reactor (HTGR) concept. Helium, the primary coolant, contains impurities (e.g., H2O and CH4) that can induce corrosion reactions at high temperatures, which in turn can affect the tribological behavior of components in sliding contact such as valves and control-rod drive systems. This paper presents results from the study of the tribological behavior of both alloys before and after conditioning them in either an impure-helium oxidizing environment or in an air environment. Both alloys were conditioned for 22 days at elevated temperatures in a once-through helium loop with 4 ppmv H2O and tested subsequently at elevated temperatures with a pin-on-disk tribometer in an air environment 650 degrees C and 750 degrees C for 800HT; 850 degrees C and 900 degrees C for 617 - with various applied loads - 1 N, 2 N and 5 N. SEM-EDS analysis revealed that conditioning the samples in an oxidizing environment leads to the formation of a mixed Fe/Cr-oxide on alloy 800HT and a Cr-oxide on alloy 617, both increasing the wear resistance compared to that of as-received samples. Alloy 617 exhibited lower steady-state friction coefficients compared to those of alloy 800HT. There was a significant decrease in the scatter of the steadystate friction coefficient of the conditioned samples compared to that of the unconditioned samples. The steady-state friction coefficient for alloy 800HT and 617 were found to be 0.53 +/- 0.07 and 0.25 +/- 0.04, respectively. The wear resistance of alloy 800HT is approximately one order of magnitude lower than that of alloy 617 in all cases that exhibited measurable wear. After sample conditioning, the wear volumes measured at low loads were undistinguishable from the unworn background, a result attributed to the formation of a compacted glaze layer under high temperatures and high contact stresses.

Alloys 800 and 617↗

High temperature oxidation and microstructural evolution of cold spray chromium coatings on Zircaloy-4 in steam environments

Oxidation kinetics and microstructural evolution of cold sprayed Cr coatings on Zircaloy-4 at 1130–1310 °C in flowing steam at atmospheric pressure have been studied. The study is aimed at understanding of the response of Cr coated Zr-alloy under a steam environment and high temperatures pertinent to design basis accidents (DBAs) and beyond design basis accidents (BDBAs) in light water reactors (LWRs). Surface morphology, microstructure, and phases of post-oxidation test samples were characterized using Scanning Electron Microscopy (SEM), x-ray diffraction (XRD), and Scanning Transmission Electron Microscopy (STEM). Growth kinetics of the Cr-oxide scale and interdiffusion layers between the Cr coating and the Zr-alloy substrate were quantified from cross-sectional SEM images. Cross-sectional analysis showed that the Cr coatings offered a 50 times reduction in oxidation rate over bare Zircaloy-4 in a 1310 °C steam environment. Oxidation kinetics at 1130 °C followed parabolic law (i.e., n ~ 0.5 in power law kinetics) but at 1230 °C and 1310 °C the value of n was suppressed to below 0.5 possibly due to the volatilization of Cr species at the two highest temperatures. Interdiffusion at coating/substrate interface resulted in formation of a brittle Cr 2 Zr or Zr(Fe, Cr) 2 intermetallic compound layer on the order of micrometer in thickness and scattered Cr-rich precipitates were observed well below the interface within the Zr-alloy substrate after cooling. These experimental results could provide data to LWR system simulation codes for better estimation of coping time in the event of accidents.

36 MATERIALS SCIENCE↗