Education

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

In-situ EBSD analysis of hydride phase transformation and its effect on micromechanical behavior in Zircaloy-4 under uniaxial tensile loading

Sun, H., Zhou, H., Luan, B., Zhang, Y., Zhu, X., Xu, C., … Liu, Q. (2024), JOURNAL OF MATERIALS RESEARCH AND TECHNOLOGY-JMR&T, 30, 6653–6667. https://doi.org/10.1016/j.jmrt.2024.05.037

Novel core-layer-shell structured nano-composite precipitates formed in selective laser melted reduced activation ferritic-martensitic steel

Yang, T., Chai, L., Tang, Y., Wu, Y., Tan, X., & Murty, K. L. (2024), MATERIALS TODAY COMMUNICATIONS, 41. https://doi.org/10.1016/j.mtcomm.2024.110242

Understanding the effect of grain boundary solute segregation on the creep behaviour of thermally stable nanocrystalline materials: a theoretical approach

Wani, I., Rai, N., Murty, K. L., & Gollapudi, S. (2024, December 16), PHILOSOPHICAL MAGAZINE. https://doi.org/10.1080/14786435.2024.2383252

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

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)

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Advancing the Technical Readiness of FeCrAl alloys and ODS Steels under Extreme Conditions for Fast Reactor Fuel Cladding

Sponsor: US Dept. of Energy (DOE)

Start Date: 10/01/22

End Date: 9/30/25

Abstract

The project team will perform several closely-knit tasks to probe the microstructural behavior and to evaluate their effects on the mechanical properties of two candidate cladding alloys ��� FeCrAl and ODS-14YWT. These alloys have been investigated for reactor use for a number of years and are widely considered to be radiation-tolerant materials that can withstand the extreme environment of a nuclear reactor. As mentioned previously, the behavior at extremely large doses and high temperatures is largely unknown. The project team, therefore, proposes ion irradiation with doses reaching to 400 dpa for temperatures ranging from 300 to 700��C on two alloys FeCrAl and ODS-14YWT. At the University of Tennessee-Knoxville (UTK) ��� Ion Beam Materials Laboratory, co-PI Weber will conduct the irradiation tests with several types of ions. These experiments will generate a database on microstructure evolution and material degradation, with irradiation temperature and high dose as key variables. Separate gas implantation effects will be probed to investigate the effects of He concentration at high irradiation doses. Thus, the effects of void interactions and void swelling, which are critical to the technical readiness of these alloys, will be evaluated at high dpas and temperatures. Leveraging the ongoing NEUP projects on miniature specimen testing, the project team will then perform in-situ thermo-mechanical experiments (tension, torsion, creep, and creep-fatigue) on the ion-irradiated samples up to a temperature of 700��C. The primary objective is to probe and characterize the microstructural changes in-situ using a scanning electron microscope (SEM). The test rig is currently being installed as a user facility by PI/PD, and co-PIs Hassan and Eapen.

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Location-Specific Material Characterization of LPBF SS316L & IN718 TCR Core Structural Materials

US Dept. of Energy (DOE)(10/01/21 - 9/30/24)

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Location-Specific Material Characterization of LPBF SS316L & IN718 TCR Core Structural Materials

Sponsor: US Dept. of Energy (DOE)

Start Date: 10/01/21

End Date: 9/30/24

Abstract

NCSU will perform impression creep and creep-fatigue tests on 316L and IN718 samples received from USU. Tests will be performed at a specified temperature and stresses around 75% of UTS and each test is expected to last less than a week. 35 to 40 tests on each of the material will be performed.

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Neutron Irradiation Data Analysis

US Dept. of Energy (DOE)(6/27/22 - 9/30/23)

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Neutron Irradiation Data Analysis

Sponsor: US Dept. of Energy (DOE)

Start Date: 6/27/22

End Date: 9/30/23

Abstract

PNNL will be getting unirradiated and neutron irradiated samples of ODS MA754 alloy. They will perform mechanical testing, FIB and TEM. FIB liftouts will be sent to CAES for APT. Thus, data collection will be performed at PNNL and CAES. NCSU (KL Murty) will be involved in data analysis and interpretation of the results.

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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)

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High Resolution Scanning Acoustic Microscopy System for High Throughput Characterization of Materials and Nuclear Fuels

Sponsor: US Dept. of Energy (DOE)

Start Date: 10/01/21

End Date: 9/30/23

Abstract

The goal of this NEUP infrastructure project is to acquire a state-of-the-art high resolution scanning acoustic microscopy system to enhance NCSU������������������s educational and research capabilities in high throughput characterization of nuclear fuels, nuclear sensor materials, cladding materials, reactor structural materials and 3D printed components.

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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)

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Novel Miniature Creep Tester for Virgin and Neutron Irradiated Clad Alloys with Benchmarked Multiscale Modeling and Simulations

Sponsor: US Dept. of Energy (DOE)

Start Date: 10/01/19

End Date: 9/30/23

Abstract

Fast and accurate measurements of creep are needed for qualifying new alloys for current and next generation reactors. For recently developed ferritic alloys such as FeCrAl, the lack of creep/fatigue data is more acute. To address this concern, this project will design and develop a novel miniature creep testing system for performing creep and load relaxation tests at multiple scales inside a scanning electron microscope (SEM). The primary objectives of the proposal are: (i) Collect rapid thermal creep and load relaxation data for two selected ferritic alloys: FeCrAl and oxide dispersed strengthened (ODS-14YWT) alloy at accelerated test conditions using solid, thin-walled and flat specimens under biaxial and uniaxial loading conditions across a temperature range of 500 ���������C to 1000 ���������C, (ii) Benchmark select data from miniature specimens against data from conventional creep tests with larger samples, (iii) Extract deformation mechanisms using in-situ SEM for virgin and neutron irradiated samples using the miniature tester, which otherwise is onerous with macroscopic creep equipment, and (iv) Perform mesoscale discrete dislocation dynamics (DD) simulations using information derived from SEM, and macroscopic constitutive modeling for predicting long-time behavior.

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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)

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Effect of Alloying and Thermo-Mechanical Processing on the Deformation of Hexagonal Close-Packed Alloys

Sponsor: National Science Foundation (NSF)

Start Date: 7/15/17

End Date: 12/31/22

Abstract

The proposal addresses the influence of thermo-mechanical-treatment (TMT) and alloying on creep anisotropy of Zircaloys that are commonly used in nuclear power reactors. The study involves characterization of anisotropic biaxial creep of Zircaloy-4 and HANA4 alloys with emphasis on the effects of recrystallization and Nb addition using closed-end internally pressurized thin-walled tubing superimposed with axial load. It is proposed here to investigate these behaviors following tube-reduction (Pilger milling), stress-relief anneal and complete recrystallization. Alloying addition of Nb has been shown to exhibit better characteristics particularly for extended use in the light water reactors with some modification of crystallographic texture and the operating slip systems and underlying creep mechanisms.

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(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)

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(20-21572) Development of an In-Situ Testing Laboratory for Research and Education of Very High Temperature Reactor Materials

Sponsor: US Dept. of Energy (DOE)

Start Date: 10/01/20

End Date: 9/30/22

Abstract

A novel thermo-mechanical fatigue (TMF) testing system, referred by miniature TMF (MTMF) system has been developed at NCSU for in-situ testing of miniature specimens within Scanning Electron Microscopes (SEM). The MTMF is capable of prescribing axial-torsional loading to solid specimen and axial-torsional-internal pressure loading to tubular specimen of 1 mm diameter at elevated temperatures (up to 1000oC) to investigate deformation of microstructure and failure mechanism in real time. Currently, in-situ SEM testing with the MTMF is performed at the Analytical Instrumentation Facility (AIF) at NCSU. This poses a serious restriction to investigate failure mechanisms of very high temperature reactor (VHTRs) materials primarily because with a user facility, such as AIF, we can only perform short-term tests that span over few days. However, fatigue, creep and creep-fatigue tests for VHTR materials may span from few days to several weeks. Hence, existing SEMs on campus are not available for long-term in-situ testing of VHTR materials. Currently, fatigue, creep and creep-fatigue failure mechanisms of new and existing alloys are mostly investigated through ex-situ testing or short duration in-situ uniaxial testing within SEM. Consequently, initiation and propagation of many failure mechanisms, especially interactions between creep and fatigue mechanisms in reducing high temperature component lives remain unknown. Hence, developing a shared in-situ testing laboratory (ISTL) is essential to allow NCSU researchers to perform novel research on nuclear materials addressing issues of fatigue, creep and creep-fatigue failure mechanisms. The proposed ISTL dedicated to performing long-term fatigue, creep and creep-fatigue tests is in critical need to develop design criteria of VHTR materials for ASME Code Sec III Div 5. However, existing facilities at NCSU or any other universities or national labs in the nation do not have a facility dedicated to perform long term tests representing realistic loading conditions of VHTR. Therefore, a suitable SEM compatible with the MTMF system at NCSU is proposed to be acquired to develop an ISTL to address high temperature nuclear materials and ASME Code issues. With the availability of such a ISTL, uniaxial and multiaxial cyclic experiments prescribing realistic thermo-mechanical fatigue (TMF), creep and creep-fatigue loading can be performed on specimens of VHTR materials, such as Alloy 617, 316H, 800H, Grade 91 steel, for addressing the high temperature component design and development issues. Finally, because of the size of commercially available TMF systems, these cannot be used for in-situ SEM testing, which is essential for investigating existing alloys and developing new alloy for VHTRs. Hence, acquisition of a SEM will give the NCSU research community unprecedented capability to perform fundamental research and educate next generation scientists in studying real-time long-term microstructure evolution of nuclear materials under uniaxial and multiaxial loading. In addition, the proposed equipment will allow training undergraduate and graduate students and postdocs in performing material characterization using advanced techniques and provide hands on experiences to students in various undergraduate and graduate courses.

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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)

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Mechanisms of Retention and Transport of Fission Products in Virgin and Irradiated Nuclear Graphite (Work Scope Identifier: RC-2)

Sponsor: US Dept. of Energy (DOE)

Start Date: 10/01/17

End Date: 9/30/21

Abstract

We propose a joint experimental-computational approach to measure the diffusivities of fission products ������������������ Iodine (I), Cesium (Cs), Krypton (Kr), Strontium (Sr), Ruthenium (Ru) and Silver (Ag), and Europium (Eu) in four graphite grades ������������������ HOPG, NBG-18, PCEA and IG-110, and uncover the mechanisms of transport using multiscale simulations involving electronic structure, atomistic, and phase field methods. By measuring the diffusivities in both virgin and neutron/ion irradiated samples (up to 25 dpa), we will probe the importance of radiation-induced defects and pores in fission product transport and retention. Non-radioactive species will be implanted into the graphite samples for simulating realistic fission product trajectories. We will leverage on our prior NEUP work on nuclear graphite, and will use ion and neutron irradiated samples from this project (NBG/HOPG/IG-110: ion irradiated, 1 and 25 dpa, PCEA: neutron irradiated, 6.61 and 10.16 dpa; in addition, the UoM will test samples from irradiated Gen I and Gen II nuclear graphite grades, some containing fission material.

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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)

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Consortium for Advanced Simulations for Light Water Reactors (CASL) - Oak Ridge National laboratory (Start up 8/16/10 to 2/17/11)

Sponsor: US Dept. of Energy (DOE)

Start Date: 11/16/10

End Date: 12/01/20

Abstract

The Consortium for Advanced Simulation of Light Water Reactors, CASL, supports the broad national missions of enabling energy independence; supporting economic growth through the offering of superior technology ; and being good stewards of the environment, buy enabling predictive simulation of nuclear power plants. Such capability will make possible power uprates, lifetime extension and higher fuel burnups for currently operating and new Generation III+ nuclear power plants.

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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)

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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

Sponsor: US Dept. of Energy (DOE)

Start Date: 10/01/16

End Date: 9/30/20

Abstract

The purpose of this proposal to develop a mechanistic understanding of accelerated fretting and wear, and bonding between Alloy 800H and Inconel 617 surfaces, we propose a series of tribological experiments in a simulated helium environment with controlled concentrations of gaseous specious, attendant microstructure characterization using electron microscopy, spectroscopy and atom-probe tomography, validated continuum models, and constitutive equations at the macroscale informed by mechanisms at the atomistic level. Our experimental data and insights derived from modeling and simulations will support ASME qualification of Alloy 800H and Inconel 617 in the new ASME Division-5 that provides construction rules for high temperature reactors and liquid metal reactors.

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