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Charit, Indrajit

Publications and source records attributed to Charit, Indrajit.

At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Pitting behavior of friction stir repair-welded 304L stainless steel in 3.5% NaCl solution at room temperature: role of grain and defect structures

Stainless steel (SS) canisters used for storing spent nuclear waste are prone to chloride-induced stress corrosion cracking. Friction stir welding (FSW), being a low heat input process, can be used for repair welding of these canisters. In this study, an artificial crack was introduced on the 304L SS coupons by an electric discharge machining process. The artificial crack was repair-welded by FSW by maintaining the tool temperature a constant at two levels (725 and 825 °C). Friction stirring significantly reduced the grain size of the stirred zone (SZ) from about 47 to 2–4 μm. The fraction of low-angle grain boundaries increased in the SZ from 2 to 37–43%. Here, on the other hand, the fraction of special grain boundaries (Σ3 and Σ9) that was ~ 50% in the base metal reduced to < 10% in the SZ. During friction stirring, the oxide layer of artificial crack was broken and aligned to form a spiral defect called lazy-S structure. All these microstructural changes affected the corrosion behavior of the FSW specimens when tested in 3.5% NaCl at room temperature using cyclic polarization, chronoamperometry, and electrochemical impedance spectroscopy. The FSW specimens showed lower polarization resistance and lower transpassive potential than those of the base metal specimens. However, the pitting protection potentials of the FSW specimens were higher than that of the base metal. The pitting behavior of the FSW specimens was influenced more by the preferential attack on the lazy-S region than by passive film breakdown. The flat band potentials of the passive film formed on the FSW specimens were more positive than that of base metal. The charge carrier density of passive film formed on the FSW specimens was higher than that of the base metal. The higher fraction of low-angle grain boundaries present in the FSW specimens could supply more number of misorientation dislocations at the metal/film interface which could form anion vacancies by a climb process leading to formation of oxide at these locations without stressing the substrate. Therefore, low-angle boundaries are considered helpful for formation of stress-free passive films that will be highly stable and enhance both pitting and stress corrosion cracking resistance.

Dry storage canister↗
DOE OSTI2020

A FEASIBILITY STUDY ON CRACK REPAIR IN AUSTENITIC STAINLESS STEEL DRY STORAGE CANISTERS USING ISOTHERMAL FRICTION STIR WELDING

In recent time, solid state crack repair techniques in spent nuclear fuel dry storage canisters (SNF-DSC) have garnered significant interest. Plates of austenitic stainless steels are usually arc welded to fabricate the cylindrical canisters which are prone to chloride-induced stress corrosion cracking. Friction stir welding (FSW) is considered to be a potential candidate for damage repair. In this work, electrical discharge machining was used to simulate cracks in 304L stainless steel plates of 12.7 mm thickness. Subsequently, isothermal FSW was carried out at two different temperatures (725 and 825°C) in order to repair the simulated crack. Microstructure-mechanical property correlations in FSWed plates were studied in detail. Significant grain refinement observed in the stir zone (SZ) was attributed to dynamic recrystallization occurring during FSW. A comparison of Vickers microhardness showed higher hardness and homogeneity in the 725°C SZ than its 825°C counterpart. The fraction of S3 boundary is found to be low in the SZ as compared to base metal (BM). Yield strength of the joints was found to be approximately 100 MPa higher than that of the BM accompanied by a drop in ductility by a factor of 2. Residual stress profile across 825°C weld was measured using x-ray diffraction.

Bhattacharyya, Madhumanti↗
DOE OSTI2020

Interactions and immobilization of lanthanides with dopants in uranium-based metallic fuels

In uranium-based fuels, production and migration of lanthanide fission products and subsequent chemical interactions with cladding constituents lead to fuel–cladding chemical interaction (FCCI), limiting fuel performance. In general, dopant addition to the fuel matrix to arrest lanthanides within the fuel by forming intermetallics is found to be effective in mitigating FCCI. Recently, we proposed ab-initio based alloy-design principles, which can be useful in identifying dopant(s) that can bind a lanthanide (Nd) inside the fuel matrix. Here, we demonstrate the robustness of such principles by performing a systematic study to choose dopants to form compounds with a range of experimentally-observed lanthanide fission products. Results reveal that for a given dopant, lanthanide Ce and La exhibit similar behavior in their compound-forming tendencies compared to Nd and Pr, which is linked to lanthanide intrinsic characteristics such as the electronic configurations. Our predicted lanthanide-dopant intermetallic formation is verified experimentally in selective cases. Finally, we showed alloy design principles that accurately identify previously known dopants like Pd, and also predict new dopants As and Se, that can be effective in binding all lanthanides within the uranium matrix. Altogether, this research helps to develop generic alloy-design principles for complex multi-component systems based on lanthanide and dopant intrinsic characteristics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Nd, SbNd and Sb 3 Nd 4 and their interactions with the cladding alloy HT9

Lanthanide fission products, such as neodymium, formed during the irradiation of metallic fuels are known to cause deleterious effects from chemical interactions occurring at the fuel-cladding interface; a phenomenon known as fuel-cladding chemical interaction (FCCI). The use of fuel-based additives that bind with the lanthanide elements within the fuel meat, alleviating their interactions at the fuel-cladding interface, is one potential method proposed to mitigate the FCCI phenomenon and extend the burnup potential of such metallic fuel systems. In this study, antimony (Sb) is evaluated as one such additive, and neodymium (Nd) is used to represent the lanthanides. A Sb-Nd alloy is fabricated, which consists of two intermetallic phases, SbNd and Sb 3 Nd 4 . Isothermal diffusion couple experiments are carried out at 675 °C for 24 h between Nd and the Fe-12Cr based HT9 cladding alloy. The results are compared against similar diffusion couple experiments carried out between the Sb-Nd alloy and also HT9. Inter-diffusion between Nd/HT9 diffusion couple is characterized and the phases Fe 17 Nd 5 and Fe 17 Nd 2 were found to form whereas no interactions were observed in the Sb-Nd/HT9 diffusion couple. The lack of compound forming tendencies between Fe, the primary alloying constituent of HT9, and SbNd is elucidated through density functional theory (DFT) calculations on enthalpy of mixing, aligning well with the experimental observations. Lastly, the strong binding of Nd with Sb appears to be favorable, alleviating Nd interactions with HT9 constituent elements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Evaluation of Tellurium as a Fuel Additive in Neodymium-Containing U-Zr Metallic Fuel

Phase-stability in a U-Zr-Te-Nd multi-component metallic fuel for advanced nuclear reactors is systematically investigated by taking into account binary, ternary and quaternary interactions between elements involved. Historically, the onset of fuel-cladding chemical interactions (FCCI) greatly limits the burnup potential of U-Zr fuels primarily due to interactions between lanthanide fission products and cladding constituents. Tellurium (Te) is evaluated as a potential additive for U-Zr fuels to bind with lanthanide fission products, e.g. neodymium (Nd), negating or mitigating the FCCI effect. Potential fresh fuel alloy compositions with the Te additive, U-Zr-Te, are characterized. Te is found to completely bind with Zr within the U-Zr matrix. Alloys simulating the formation of the lanthanide element Nd within U-Zr-Te are also evaluated, where the Te-Nd binary interaction dominates and NdTe is found to form as a high temperature stable compound. The experimental observations agree well with the trends obtained from density functional theory calculations. According to the calculated enthalpy of mixing, Zr-Te compound formation is favored in the U-Zr-Te alloy whereas NdTe compound formation is favored in the U-Zr-Te-Nd alloy. Further, the calculated charge density distribution and density of states provide sound understanding of the mutual chemical interactions between elements and phase-stability within the multi-component fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Creep behavior of Alloy 709 at 700 °C

Alloy 709 is an austenitic stainless steel with potential for structural applications in sodium-cooled fast reactors. In this study, creep tests on Alloy 709 specimens were performed at a temperature of 700°C under applied stresses in the range of 125 to 250 MPa. Here, the alloy exhibited a brief primary, insignificant secondary and prolonged tertiary creep stages. A Norton stress exponent (n) of ~6 was obtained from the creep plots. Microstructural analyses were carried out for both as-received and crept specimens using optical microscopy, electron backscattered diffraction, and transmission electron microscopy. The microstructures of the as-received and crept specimens were compared. The mode of fracture was transgranular as revealed by scanning electron microscopy of the creep-ruptured specimens. Thermodynamic and precipitation modeling was performed in order to obtain information on the phase types and the kinetics of precipitation for the sigma-phase and the M 23 C 6 carbide. Based on the creep data, microstructural observations and relevant modeling, creep deformation and failure mechanisms at the considered temperature are elucidated.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Development of Fe-9Cr Alloy via High-Energy Ball Milling and Spark Plasma Sintering

Iron and chromium powders were mixed and mechanically alloyed via high-energy ball milling for different time durations (2–20 h) to produce an alloy powder with a nominal composition of Fe-9Cr (wt.%). The milled Fe-9Cr powders were analyzed using scanning electron microscopy (SEM), x-ray diffraction (XRD) and transmission electron microscopy (TEM) to understand the impact of ball milling (BM) time on the characteristics of the milled powder. The optimized milled powder samples were then consolidated via spark plasma sintering (SPS) for different dwell times and temperatures. The density of consolidated samples was found to reach a maximum of 98%. Microstructural characterization of the SPS samples were performed using XRD, SEM, TEM and electron backscatter diffraction. This study highlights the principles and importance of high-energy BM and SPS of Fe-9Cr model alloy for the future development of more complex oxide dispersion-strengthened alloys for various applications including advanced nuclear reactor applications.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Characterization of Creep-Fatigue Crack Growth in Alloy 709 and Prediction of Service Lives in Nuclear Reactor Components (Final Report)

This final report summarizes the activities and results that were obtained by the research team working on the project “Characterization of Creep-Fatigue Crack Growth in Alloy 709 and Prediction of Service Lives in Nuclear Reactor Components”, DOE-NEUP Project 15-8623. This project was mainly concerned with the creep-fatigue crack growth characterization of austenitic stainless-steel Fe-20Cr-25Ni (Alloy 709). For this purpose, an array of experimental and computational activities was performed. The main goal of these activities was to describe and predict the crack growth behavior in Alloy 709 under different conditions of applied loading and temperature. Alloy 709 specimens were tested to describe both the uniaxial creep behavior and creep-fatigue crack growth rates at temperatures between 550°C and 700°C. Microstructure and fracture surface analysis was performed using several experimental techniques, and the defining features of crack growth propagation were identified and described. Thermodynamic and diffusion simulations were performed to understand the microstructure evolution of Alloy 709 at elevated temperatures, and to establish an aging schedule relevant for in-service conditions over extended periods. Experimental testing was performed on both as-received and aged Alloy 709 specimens. Two computational methods were developed to predict crack growth rates and remaining lives in specimens loaded under either fatigue of creep-fatigue. The first computational technique was the finite element method, while the second was strip-yield modeling. Both methods relied on the prediction of crack growth rates using the concept of plasticity-induced crack closure. Crack-tip opening loads were computed and used to perform crack growth rate predictions. The computational predictions matched well with the experimental results. The strip-yield model was shown to be an excellent predictive tool not only for crack growth in austenitic SS Alloy 709, but also in 316 SS, martensitic Cr-Mo steels, Ni-based alloys and Al-based alloys. The main conclusion of this study is that Alloy 709 exhibits excellent creep crack growth resistance at temperatures between 550°C and 700°C. In fact, the crack growth resistance of this alloy appears to be higher than that of 316 SS in similar testing conditions. Moreover, when creep loading with hold (dwell) times of 60s or 600s were intercalated every fatigue cycle, the crack growth rates did not change significantly. The conclusion is that the tests performed revealed that the crack growth is insignificant during the creep portion of a loading cycle, as compared with the crack growth during the fatigue portion of the cycle. In general, Alloy 709 shows excellent resistance to creep deformations and creep crack growth up to 700°C.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Characterization of U-10Zr-2Sn-2Sb and U-10Zr-2Sn-2Sb-4Ln to assess Sn+Sb as a mixed additive system to bind lanthanides

In this paper, tin and antimony are being investigated as potential additives to metallic fuel to control fuel-cladding chemical interaction (FCCI). A primary cause of FCCI is lanthanide fission products moving to the fuel periphery and interacting with the cladding. This interaction can lead to wastage of the cladding and, given enough time or burn-up, eventually to a cladding breach. The current study involves a microstructural characterization of as-cast and annealed U-10Zr-2Sn-2Sb and U-10Zr-2Sn-2Sb-4Ln, where Ln = 53Nd-25Ce-16Pr-6La, all in wt %, as alloys which could prevent FCCI. Scanning electron microscopy (SEM) analysis and X-ray powder diffraction indicates the additive-Zr compound in U-10Zr-2Sn-2Sb is Zr 5 (Sn,Sb) 3 , while the additive-Ln compound being formed in U-10Zr-2Sn-2Sb-4Ln is Ln 5 (Sn,Sb) 4 , with significantly more Sb than Sn (37 at. % versus 6 at. %, respectively). The bulk of the Sn, with a small amount of Sb, remained as Zr 5 (Sn,Sb) 3 precipitates. Finally, the potential benefits of a mixed additive system, along with relative stabilities of the intermetallic compounds, are discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Transmission electron microscopy of ion irradiated ODS MA956 samples

Oxide dispersion strengthened (ODS) alloys are promising candidate materials for the next generation advanced nuclear reactors due to their superior irradiation resistance and mechanical properties. To better understand the effect of irradiation on MA956, it is essential to study higher dose (50-100 dpa) samples, so that the general trend of microstructural evolution and the resulting radiation-hardening can be deduced. Currently, ion irradiations are considered the only way to achieve doses beyond ~50 dpa in a practical time frame relevant to alloy and welding development programs. This dataset contains TEM characterization results of ODS MA956 samples that were ion irradiated under different conditions: Sample #5 (1.25 dpa at 190℃); Sample #9 (50 dpa at 190℃); Sample #15 (1.25dpa under 320℃); and Sample 17 (25 dpa at 320℃). TEM characterization focused on the irradiation induced defects (dislocation lines and loops) using the on-zone axis bright field STEM technique. These data were collected using a FEI Tecnai G2 F30 S/TEM at Microscopy and Characterization Suite (MaCS) at Center for Advanced Energy Studies (CAES), Idaho Falls, ID. This project (ion irradiation and TEM studies) was supported by the U.S. Department of Energy, Office of Nuclear Energy under DOE Idaho Operations Office Contract DE-AC07-05ID14517 as part of Nuclear Science User Facilities award #18-14784 (PI: Ramprashad Prabhakaran, PNNL).

Prabhakaran, Ramprashad↗
DOE OSTI2018

Formability of a wrought Mg alloy evaluated by impression testing

This study is focused on furthering our understanding of the different factors that influence the formability of Magnesium alloys. Towards this end, formability studies were undertaken on a wrought Mg-2Zn-1Mn (ZM21) alloy. In contrast to conventional formability studies, the impression testing method was adopted here to evaluate the formability parameter, B, at temperatures ranging from 298 to 473 K. The variation of B of ZM21 with temperature and its rather limited values were discussed in the light of different deformation mechanisms such as activation of twinning, slip, grain boundary sliding (GBS) and dynamic recrystallization (DRX). It was found that the material characteristics such as grain size, texture and testing conditions such as temperature and strain rate, were key determinants of the mechanism of plastic deformation. A by-product of this analysis was the observation of an interesting correlation between the Zener-Hollomon parameter, Z, and the ability of Mg alloys to undergo DRX.

Formability↗
DOE OSTI2017

Integral Inherently Safe Light Water Reactor (I 2 S-LWR)

This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I 2 S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Boride Based Electrode Materials with Enhanced Stability Under Extreme Conditions for MHD Direct Power Extraction. Final Scientific Report

Coal power currently makes up almost 40% of global electricity production. This method of extracting energy from coal can be added to any coal plant, improving its energy output. Not only would this prepare us to keep up with the ever-increasing energy demands but it could also allow us to significantly cut the greenhouse gas emissions for the coal power industry. Hafnium diboride (HfB 2 ), Zirconium diboride (ZrB 2 ), and their solid solution mixtures were studied, primarily for electrode coating applications in coal-fired magnetohydrodynamic (MHD) direct power extraction plants. MHD works by directly extracting current from superheated plasma flame (2725 °C) seeded with potassium salts. Obviously, the electrodes near this flame are subject to extreme temperatures and corrosive environments. The coating material needs to protect the electrodes from salt corrosion, oxidation, and melting, while still providing electrical conductivity. Transition metal diborides have many of the desired properties but need improvement for oxidation resistance above 1200 °C. Two main methods of oxidation resistance improvements were investigated during this research. The first was through electrochemical anodization, a process where a thin protective oxide layer is grown on the surface to prevent further detrimental oxidation at high temperatures. Several solutions, voltages, and methods were used to find an oxide layer that would protect at high temperatures. The second method was to use various additives and mixture ratios to improve densification and oxidation resistance. Ratios of HfB 2 and ZrB 2 tested were 1:4, 1:1, 4:1 and 1:1 with additives of Ta, Hf, and Zr metals, bringing the boron to metal ratio down to 1.86. Additionally, 1.8 mol% Gd 2 O 3 and LaB 6 were added to the 1:1 ratio HfB 2 -ZrB 2 . Hydrogen evolution reaction kinetics were also investigated to shed some light on the ability of these materials to resist corrosion. Aqueous corrosion tests were also performed in several solutions of different pH. The 1:1 ratio solution added with LaB 6 gave the best overall results both in high temperature and room temperature aqueous experiments. Further work will be needed to evaluate the efficacy of these materials for such applications.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2017

FY17 CAES LDRD Annual Report

Metallic alloys are widely used or planned for use as structural and cladding materials in current and future reactors. Under irradiation, grain boundary (GB) cohesion strength decreases due to interaction with defects and impurities, leading to intergranular fracture and embrittlement of alloys. The objective of this project is to develop a technique for quantifying GB cohesion and its impact on fracture behavior in irradiated alloys, by utilizing transmission electron microscopic (TEM) in situ cantilever testing in concert with multi-scale modeling. The TEM in situ cantilever testing is a novel approach for studying the real-time mechanical response of materials. It will be used in this work for studying intergranular fracture behavior in several irradiated iron-based ferritic alloys and providing key information to link atomistic level events with mesoscale/macroscopic mechanical properties. The Multi-Physics Object-Oriented Simulation Environment (MOOSE)-based cohesive zone model (CZM) and extended finite element method (XFEM) for intergranular fracture of irradiated ferritic alloys will be developed in this work by utilizing atomistic results as inputs and experimental results for validation.

36 MATERIALS SCIENCE↗
DOE OSTI2016

X-ray absorption spectroscopy characterization of embedded and extracted nano-oxides

Here, the chemistries and structures of both embedded and extracted Ysingle bondTisingle bondO nanometer-scale oxides in a nanostructured ferritic alloy (NFA) were probed by x-ray absorption spectroscopy (XAS). Y 2 Ti 2 O 7 is the primary embedded phase, while the slightly larger extracted oxides are primarily Y 2 TiO 5 . Analysis of the embedded nano-oxides is difficult partly due to the multiple Ti environments associated with different oxides and those still residing in matrix lattice sites. Thus, bulk extraction followed by selective filtration was used to isolate the larger Y 2 TiO 5 oxides for XAS, while the smaller predominant embedded phase Y 2 Ti 2 O 7 oxides passed through the filters and were analyzed using the log-ratio method.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗