Korukonda Murty

Progress Energy Distinguished Professor of Nuclear Engineering

  • 919-515-3657
  • 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

Ph.D. 1967

Materials Science

Cornell University

M.S. 1967

Materials Science

Cornell University

M.S. 1963

Physics

Andhra University

B.S. 1962

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

Exploring the stability, thermodynamic and mechanical properties of zirconium oxides and suboxides under temperature and pressure: A first-principles predictions
Zhou, H., Luan, B., Chen, L., Yang, X., Liu, C., Liu, X., … Murty, K. L. (2024), JOURNAL OF NUCLEAR MATERIALS, 591. https://doi.org/10.1016/j.jnucmat.2024.154934
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
Strength-ductility synergy through tailoring heterostructures of hot-rolled ferritic-martensitic steels containing varying Si contents
Yang, T., Chai, L., Wang, H., Li, G., & Murty, K. L. (2023), MATERIALS SCIENCE AND ENGINEERING A-STRUCTURAL MATERIALS PROPERTIES MICROSTRUCTURE AND PROCESSING, 886. https://doi.org/10.1016/j.msea.2023.145712
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, Vol. 6. 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

View all publications via NC State Libraries

Grants

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 - 9/30/24)
Neutron Irradiation Data Analysis
US Dept. of Energy (DOE)(6/27/22 - 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)
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)
Effect of Alloying and Thermo-Mechanical Processing on the Deformation of Hexagonal Close-Packed Alloys
National Science Foundation (NSF)(7/15/17 - 12/31/22)
(20-21572) 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)
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)
Consortium for Advanced Simulations for Light Water Reactors (CASL) - Oak Ridge National laboratory (Start up 8/16/10 to 2/17/11)
US Dept. of Energy (DOE)(11/16/10 - 12/01/20)
Tribological Damage Mechanisms from Experiments and Validated Simulations of Alloy 800H and Inconel 617 in a Simulated HTGR/VHTR Helium Environment, RC-2.3: Helium Tribology for HTGR's
US Dept. of Energy (DOE)(10/01/16 - 9/30/20)