Korukonda Murty
Progress Energy Distinguished Professor of Nuclear Engineering

- 919-515-3657
- murty@eos.ncsu.edu
- Burlington Laboratory 3143
I have been interested in the deformation, creep, fatigue and fracture behaviors of nuclear core and pressure boundary materials, with particular emphasis on structureproperty relationship and effects of radiation exposure. I am also interested in radiation-enhanced hydrogen transport into steels used for radioactive waste containers and the subsequent embrittlement with reference to their integrity. We are actively pursuing studies on the effects of fabrication processes on crystallographic texture and the resulting anisotropic mechanical properties of Zircaloy cladding, with application to the understanding of pellet-cladding mechanical interaction.
One of the major areas of my research is on the basic deformation mechanisms in materials, and I have been developing model creep equations to predict the in-service materials’ deformation behavior. In addition, I am interested in the micro mechanisms of macroscopic crack propagation. These studies involve mechanical testing, sub structural studies using electron microscopy and basic deformation model development. One of the research areas underway is the effect of crystallographic texture, stacking-fault energy and crystal structure on the anisotropic mechanical deformation and creep of hexagonal close-packed metals.
Other research areas include in-site NMR studies of the dynamical behavior of point and line defects in materials during deformation, ball indentation for non-destructive evaluation of materials’ condition in-service and mechanical integrity of electronic packaging materials.
Education
Materials Science
Cornell University
Materials Science
Cornell University
Physics
Andhra University
Physics
Andhra University
Research Description
Dr. Murty is interested in the deformation, creep, fatigue and fracture behaviors of nuclear core and pressure boundary materials, with particular emphasis on structureproperty relationship and effects of radiation exposure.
Publications
- Microstructures and wear resistance of Zr-4 and N36 alloys subjected to pulsed laser surface remelting
- Zhang, F., Chai, L., Qi, L., Wang, Y., Wu, L., Pan, H., … Murty, K. L. (2023), JOURNAL OF NUCLEAR MATERIALS, 577. https://doi.org/10.1016/j.jnucmat.2023.154284
- A quasi in-situ study on the work hardening and softening mechanisms of Ti-33Zr-12Al-6V alloy
- Zhang, F., Luan, B., Shou, H., Zheng, J., Zhang, X., Liu, R., & Murty, K. L. (2022), MATERIALS SCIENCE AND ENGINEERING A-STRUCTURAL MATERIALS PROPERTIES MICROSTRUCTURE AND PROCESSING, 835. https://doi.org/10.1016/j.msea.2022.142694
- High Temperature Deformation Behavior of a Fe-25Ni-20Cr (Wt Pct) Austenitic Stainless Steel
- Alomari, A. S., Kumar, N., Hawary, M., & Murty, K. L. (2022, June 14), METALLURGICAL AND MATERIALS TRANSACTIONS A-PHYSICAL METALLURGY AND MATERIALS SCIENCE. https://doi.org/10.1007/s11661-022-06739-6
- Revealing Microstructural, Textural, and Hardness Evolution of Ti-6Al-4V Sheet Cooled From Sub beta-Transus Temperature at Different Rates
- Chai, L., Xia, J., Murty, K. L., Gu, X., Fan, J., & Yao, Z. (2022, June 14), METALLURGICAL AND MATERIALS TRANSACTIONS A-PHYSICAL METALLURGY AND MATERIALS SCIENCE. https://doi.org/10.1007/s11661-022-06737-8
- A strategy to introduce gradient equiaxed grains into Zr sheet by combining laser surface treatment, rolling and annealing
- Chai, L., Zhu, Y., Hu, X., Murty, K. L., Guo, N., Chen, L.-Y., … Zhang, L.-C. (2021), SCRIPTA MATERIALIA, 196. https://doi.org/10.1016/j.scriptamat.2021.113761
- Effect of Strain Range on High Temperature Creep-Fatigue Behaviour of Fe-25Ni-20Cr (wt.%) Austenitic Stainless Steel (Alloy 709)
- Alsmadi, Z. Y., & Murty, K. L. (2021), MATERIALS AT HIGH TEMPERATURES, 38(1), 47–60. https://doi.org/10.1080/09603409.2020.1859310
- Effect of friction stir processing and subsequent annealing on microstructure and mechanical properties of a metastable beta-Zr alloy
- Li, S., Luan, B., Liao, Z., Liu, Z., Chu, L., Wen, S., … Liu, Q. (2021), MATERIALS SCIENCE AND ENGINEERING A-STRUCTURAL MATERIALS PROPERTIES MICROSTRUCTURE AND PROCESSING, 822. https://doi.org/10.1016/j.msea.2021.141660
- Effects of composition on phase stabilities and elastic properties in TiZrAlV alloys : Experiments and first-principles calculations
- Zhang, F., Luan, B., Chu, L., Wen, S., Zhang, S., Wang, Y., … Murty, K. L. (2021), JOURNAL OF ALLOYS AND COMPOUNDS, 863. https://doi.org/10.1016/j.jallcom.2020.158054
- High-temperature effects on creep-fatigue interaction of the Alloy 709 austenitic stainless steel
- Alsmadi, Z. Y., & Murty, K. L. (2021), INTERNATIONAL JOURNAL OF FATIGUE, 143. https://doi.org/10.1016/j.ijfatigue.2020.105987
- A multi-deformation mechanisms assisted metastable beta-ZrTiAlV alloy exhibits high yield strength and high work hardening rate
- Liao, Z., Luan, B., Zhang, X., Liu, R., Murty, K. L., & Liu, Q. (2020), JOURNAL OF ALLOYS AND COMPOUNDS, 816. https://doi.org/10.1016/j.jallcom.2019.152642
Grants
- Effect of neutron irradiation on friction stir welded Ni-based ODS MA754 alloy NSUF 1.2 (Irradiation testing of materials produced by innovative manufacturing techniques)
- US Dept. of Energy (DOE)(6/27/22 - 9/30/23)
- Advancing the Technical Readiness of FeCrAl alloys and ODS Steels under Extreme Conditions for Fast Reactor Fuel Cladding
- US Dept. of Energy (DOE)(10/01/22 - 9/30/25)
- Location-Specific Material Characterization of LPBF SS316L & IN718 TCR Core Structural Materials
- US Dept. of Energy (DOE)(10/01/21 - 9/30/23)
- High Resolution Scanning Acoustic Microscopy System for High Throughput Characterization of Materials and Nuclear Fuels
- US Dept. of Energy (DOE)(10/01/21 - 9/30/23)
- Development of an In-Situ Testing Laboratory for Research and Education of Very High Temperature Reactor Materials
- US Dept. of Energy (DOE)(10/01/20 - 9/30/22)
- Novel Miniature Creep Tester for Virgin and Neutron Irradiated Clad Alloys with Benchmarked Multiscale Modeling and Simulations
- US Dept. of Energy (DOE)(10/01/19 - 9/30/23)
- Mechanisms of Retention and Transport of Fission Products in Virgin and Irradiated Nuclear Graphite. Work Scope Identifier: RC-2.
- US Dept. of Energy (DOE)(10/01/17 - 9/30/21)
- Effect of alloying and thermo-mechanical processing on biaxial creep of low c/a-ratio hexagonal closed packed metals (Zr-alloys)
- National Science Foundation (NSF)(7/15/17 - 12/31/22)
- Advanced Nuclear Materials Laboratory Enhancements for Corrosion and Stress Corrosion Testing
- US Dept. of Energy (DOE)(10/01/17 - 9/30/19)
- Tribological Damage Mechanisms from Experiments and Validated Simulations of Alloy 800H and Inconel 617 in a Simulated HTGR/VHTR Helium Environment
- US Dept. of Energy (DOE)(10/01/16 - 9/30/20)