Progress Energy Distinguished Professor of Nuclear Engineering, Director of Nuclear Engineering Graduate Program
- 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.
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.
- alpha ->beta transformation characteristics revealed by pulsed laser-induced non-equilibrium microstructures in duplex-phase Zr alloy
- Chai, L. J., Wang, S. Y., Wu, H., Guo, N., Pan, H. C., Chen, L. Y., Murty, K. L., & Song, B. (2017), Science China-Technological Sciences, 60(8), 1255-1262.
- Fracture behavior and grain boundary sliding during high-temperature low-stress deformation of AZ31 magnesium alloy
- Roodposhti, P. S., & Murty, K. L. (2017), In Mechanical and creep behavior of advanced materials. (Minerals Metals & Materials Series, ) (pp. 279-287).
- Effect of Mo and Bi additions on the microstructure of Zr-Cr-Fe alloy after beta-quenching
- Wang, J. M., Luan, B. F., Murty, K. L., & Liu, Q. (2017), In Mechanical and creep behavior of advanced materials. (Minerals Metals & Materials Series, ) (pp. 183-192).
- Dislocation cross-slip controlled creep at high stresses and transitional creep mechanisms in zircaloy-4
- Kombaiah, B., & Murty, K. L. (2017), In Mechanical and creep behavior of advanced materials. (Minerals Metals & Materials Series, ) (pp. 65-77).
- Microstructural, textural and hardness evolution of commercially pure Zr surface-treated by high current pulsed electron beam
- Chai, L. J., Chen, B. F., Wang, S. Y., Zhang, Z., & Murty, K. L. (2016), Applied Surface Science, 390, 430-434.
- Effects of microstructure and processing methods on creep behavior of AZ91 magnesium alloy
- Roodposhti, P. S., Sarkar, A., Murty, K. L., & Scattergood, R. O. (2016), Journal of Materials Engineering and Performance, 25(9), 3697-3709.
- Grain boundary sliding mechanism during high temperature deformation of AZ31 Magnesium alloy
- Roodposhti, P. S., Sarkar, A., Murty, K. L., Brody, H., & Scattergood, R. (2016), Materials Science & Engineering. A, Structural Materials: Properties, Microstructure and Processing, 669, 171-177.
- The role of grain size on neutron irradiation response of nanocrystalline copper
- Mohamed, W., Miller, B., Porter, D., & Murty, K. (2016), Materials, 9(3).
- Modeling irradiation creep of graphite using rate theory
- Sarkar, A., Eapen, J., Raj, A., Murty, K. L., & Burchell, T. D. (2016), Journal of Nuclear Materials, 473, 197-205.
- Microstructural and textural evolution of commercially pure Zr sheet rolled at room and liquid nitrogen temperatures
- Chai, L. J., Luan, B. F., Xiao, D. P., Zhang, M., Murty, K. L., & Liu, Q. (2015), Materials & Design, 85, 296-308.
- North Carolina State Universityâ€™s Graduate Fellowship in Nuclear Engineering (NCSUâ€“GFINE) - 2016
- US Nuclear Regulatory Commission(7/01/16 - 6/30/20)
- National Academy for Nuclear Training Fellowship Program 2015-2016
- Institute of Nuclear Power Operations (INPO)(1/01/16 - 12/31/16)
- Innovative Approach to SCC Inspection and Evaluation of Canister in Dry Storage
- US Dept. of Energy (DOE)(10/01/15 - 9/30/16)
- Mechanistic and Validated Creep/Fatigue Predictions for Alloy 709 from Accelerated Experiments and Simulations
- US Dept. of Energy (DOE)(10/01/15 - 9/30/18)
- National Academy for Nuclear Training Fellowship Program 2014-2015
- National Academy for Nuclear Training(1/01/15 - 12/31/15)
- National Academy for Nuclear Training Fellowship Program 2013-2014
- National Academy for Nuclear Training(1/01/14 - 12/31/14)
- Study on High Temperature Creep Behavior of Zirconium Alloys
- Korea Atomic Energy Research Institute (KAERI)(9/01/13 - 8/31/15)
- MRI: Development of a Miniature, High Temperature, Multiaxial Testing System for Advanced Materials and Engineering Research
- National Science Foundation (NSF)(8/15/13 - 7/31/17)
- North Carolina State University's Graduate Fellowship In Nuclear Engineering (NCSU-GFINE)
- US Nuclear Regulatory Commission(8/01/13 - 7/31/17)
- Aging of Used Nuclear Fuel in Storage: Accelerated Characterization and Modeling of Performance
- US Dept. of Energy (DOE)(12/15/11 - 12/15/15)