Priyanka Agrawal
- Name
- Dr. Priyanka Agrawal
- Institution
- University of North Texas
- Position
- Research Assistant Professor
- Affiliation
- University of North Texas
- h-Index
- 17
- ORCID
- 0000-0001-9894-9625
- Biography
Dr. Priyanka Agrawal is a Research Assistant Professor at the University of North Texas, Denton, in the Materials Science and Engineering department. Before this, she was a Post Doctorate Research Associate at Ames National Laboratory and the University of North Texas, Denton. She did her Ph.D. at the Department of Materials Engineering, Indian Institute of Science, INDIA involving designing and setting up Creep and Fatigue testing machines, carrying out mechanical tests, and subsequent Electron Microscopy to understand and relate mechanical properties to the aspects of microstructure and model the phenomenon. Dr. Agrawal is an expert in the mechanical behavior of materials and transmission electron microscopy (>15yrs), working on materials used in automobile, aerospace, ballistic, and nuclear applications. She is proficient in evaluating the deformation mechanisms for advanced manufacturing processes like additive manufacturing and friction stir processing and its derivatives like additive friction stir deposition (AFSD) and SSE, via microstructure and property correlation and analytical modeling, as reflected in her ~45 publications in high-impact journals and is a Reviewer for many journals.
- Expertise
- Additive Manufacturing, Friction Stir Welding, High Entropy Alloy (HEA), Titanium, Transmission Electron Microscopy (TEM)
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"Ion irradiation and examination of Additive friction stir deposited 316 stainless steel"
Priyanka Agrawal, Ching-Heng Shiau, Aishani Sharma, Zhihan Hu, Megha Dubey, Yu Lu, Lin Shao, Ramprashad Prabhakaran, Yaqiao Wu, Rajiv Mishra,
Materials & Design
Vol. 238
2024
112730
Link
This study explored solid-state additive friction stir deposition (AFSD) as a modular manufacturing technology, with the aim of enabling a more rapid and streamlined on-site fabrication process for large meter-scale nuclear structural components with fully dense parts. Austenitic 316 stainless steel (SS) is an excellent candidate to demonstrate AFSD, as it is a commonly-used structural material for nuclear applications. The microstructural evolution and concomitant changes in mechanical properties after 5 MeV Fe++ ion irradiation were studied comprehensively via transmission electron microscopy and nanoindentation. AFSD-processed 316 SS led to a fine-grained and ultrafine-grained microstructure that resulted in a simultaneous increase in strength, ductility, toughness, irradiation resistance, and corrosion resistance. The AFSD samples did not exhibit voids even at 100 dpa dose at 600 °C. The enhanced radiation tolerance as compared to conventional SS was reasoned to be due to the high density of grain boundaries that act as irradiation-induced defect sinks. |
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"Irradiation response of innovatively engineered metastable TRIP high entropy alloy"
Priyanka Agrawal,
Journal of Nuclear Materials
Vol. 574
2022
Link
Properties and radiation responses of a metastable high entropy alloy (HEA) exhibiting the transformation induced plasticity (TRIP) effect were studied. Innovative engineering used to manufacture this HEA has shown superior mechanical and corrosion properties in 3.5% NaCl than most advanced stainless steels. The microstructural evolution and corresponding mechanical response after irradiation have been evaluated using detailed transmission electron microscopy, and nanoindentation. The study shows a change in metastability of the alloy with irradiation via a recovery mechanism, where the irradiation-induced transformation is reversed by the temperature-induced transformation, thereby introducing the concept of self-healing, made possible due to the TRIP behavior of HEA. |
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"Irradiation-induced shift in the thermodynamic stability of phases and the self-healing effect in transformative high entropy alloys "
Priyanka Agrawal,
Journal of Nuclear Materials
Vol. 597
2024
Link
This work investigated Fe40Mn20Cr15Co20Si5 high entropy alloy (CS-HEA), which exhibits transformation induced plasticity (TRIP) from -fcc-hcp, as a probable candidate for nuclear applications. CS-HEA is an extensively-explored, low stacking fault energy alloy with superior strength, ductility, fatigue resistance, and corrosion resistance. This study delved into the effect of irradiation on the shift in thermodynamic stability of the phases and thus radiation tolerance. The evolution of phases, lattice parameters, and transformation volume, V_(γ→ε,) were evaluated from X-ray diffraction experiments along with the mechanical response from nanoindentation. The alloy exhibited a recently-proposed novel self-healing mechanism possible due to the TRIP effect to minimize irradiation damage by restraining the -fcc-hcp transformation via thermal aid; this self-healing mechanism was confirmed by transmission electron microscopy. The results were corroborated by a negative change in V_(γ→ε) and a low |V_(γ→ε) |, which is an important criterion for recovery of parent -fcc phase. Thus, this alloy was deemed a good radiation-tolerant candidate for nuclear application. |
| "Ion Irradiation and Examination of Additive Friction Stir Manufactured 316 Stainless Steel Component" Priyanka Agrawal, TMS 2024 March 3-7, (2024) Link |
About Us
The Nuclear Science User Facilities (NSUF) is the U.S. Department of Energy Office of Nuclear Energy's only designated nuclear energy user facility. Through peer-reviewed proposal processes, the NSUF provides researchers access to neutron, ion, and gamma irradiations, post-irradiation examination and beamline capabilities at Idaho National Laboratory and a diverse mix of university, national laboratory and industry partner institutions.
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