Education

Research Description

Dr. Eapen's research interests are in science-based modeling and simulation of nuclear fuel for advanced reactors, nano-structured materials, plasma-material interactions, energy transport at nano-scales and interfaces, modeling of diffusion and viscosity in glass matrices for radioactive waste management.

Publications

Corrosion resistance and mechanical properties of Armakap cement for nuclear applications

Moneghan, D., Shkoukani, G., Eapen, J., Bourham, M., Mustafa, W., & Kilani, R. (2021), NUCLEAR ENGINEERING AND DESIGN, 374. https://doi.org/10.1016/j.nucengdes.2021.111049

Deducing Phonon Scattering from Normal Mode Excitations

Raj, A., & Eapen, J. (2019), SCIENTIFIC REPORTS, 9. https://doi.org/10.1038/s41598-019-43306-3

Exact diagonal representation of normal mode energy, occupation number, and heat current for phonon-dominated thermal transport

Raj, A., & Eapen, J. (2019), JOURNAL OF CHEMICAL PHYSICS, 151(10). https://doi.org/10.1063/1.5099936

Phonon dispersion using the ratio of zero-time correlations among conjugate variables: Computing full phonon dispersion surface of graphene

Raj, A., & Eapen, J. (2019), COMPUTER PHYSICS COMMUNICATIONS, 238, 124–137. https://doi.org/10.1016/j.cpc.2018.12.008

Approach to local thermodynamic equilibrium and the evolution to a glassy core following neutron/ion radiation impact

Mei, X., Mohamed, W., & Eapen, J. (2018), PHILOSOPHICAL MAGAZINE, 98(29), 2701–2722. https://doi.org/10.1080/14786435.2018.1502482

Computing phonon dispersion using fast zero-point correlations of conjugate variables

Raj, A., & Eapen, J. (2018), MRS Advances, 3(10), 531–536. https://doi.org/10.1557/adv.2018.288

Thermal jamming of ions in the superionic state of UO2

Sanders, D., & Eapen, J. (2018), MRS Advances, 3(31), 1777–1781. https://doi.org/10.1557/adv.2018.326

Using space-time correlations to identify transient defects

Lowe, W., & Eapen, J. (2018), MRS Advances, 3(31), 1755–1760. https://doi.org/10.1557/adv.2018.196

Disordering and dynamic self-organization in stoichiometric UO2 at high temperatures

Annamareddy, A., & Eapen, J. (2017), Journal of Nuclear Materials, 483, 132–141. https://doi.org/10.1016/j.jnucmat.2016.10.042

Entropic crossovers in superionic fluorites from specific heat

Eapen, J., & Annamareddy, A. (2017), Ionics, 23(4), 1043–1047. https://doi.org/10.1007/s11581-017-2007-z

View all publications via NC State Libraries

Grants

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 - 9/30/21)

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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: 9/30/21

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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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 - 9/30/22)

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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: 9/30/22

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 ºC to 1000 ºC, (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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Evaluation of the Performance of ArmaKap Ceramic Cement as a Radiation Shielding Materials with Experiments and Simulations

ArmaKap Technologies(9/01/17 - 12/31/20)

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Evaluation of the Performance of ArmaKap Ceramic Cement as a Radiation Shielding Materials with Experiments and Simulations

Sponsor: ArmaKap Technologies

Start Date: 9/01/17

End Date: 12/31/20

Abstract

ArmaKap Technologies developed Ceramic Cement as an innovative material that can be used for radiation shielding. Preliminary results on such materials have shown better attenuation of gamma rays as compared to conventional concrete mixes. Exposure to radiation comes from various sources such as cosmic rays, highly energetic radiation from outer space and terrestrial natural radiation. Natural radiation included naturally-radioactive elements. Additional radiation sources x-rays in medical facilities, nuclear reactors, nuclear weapons, cathode ray tubes used in TV and computer displays, and numerous other radiation-producing devices. Magnitude of the radiation dose in any radiation-producing facility, as well as natural sources, must be controlled to eliminate exposure to radiation, or to limit exposure to the regulatory standards. This proposal aims to conduct research on the ArmaKap ceramic cement to determine its gamma ray attenuation efficiency and its appropriateness for radiation shielding when used in nuclear and waste disposal facilities.

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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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Advanced Nuclear Materials Laboratory Enhancements for Corrosion and Stress Corrosion Testing

US Dept. of Energy (DOE)(10/01/17 - 9/30/19)

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Advanced Nuclear Materials Laboratory Enhancements for Corrosion and Stress Corrosion Testing

Sponsor: US Dept. of Energy (DOE)

Start Date: 10/01/17

End Date: 9/30/19

Abstract

Funding is requested, through the DOE NEUP Infrastructure program, to obtain the acquisition of autoclaves and equipment for Corrosion, Stress-assisted Corrosion, and Stress Corrosion Cracking testing of nuclear materials for structural and cladding applications in LWRs, and corrosion studies related to HLW storage packages. The proposed equipment and laboratory enhancements will be used in the performance of research related to existing DOE projects and proposed efforts in the areas of advanced nuclear fuels and materials. The equipment will also be used for the purpose of a laboratory class which is being developed by the PIs for inclusion in the curriculum. The 1 credit Lab is in complement to the 3 credit Nuclear Materials Science class. It will also be of relevance for a class on Corrosion of Nuclear Materials developed by two of the co-PIs at the graduate level which will cover issues related to corrosion and stress corrosion of materials used in nuclear reactors. Henceforward, the requested infrastructure will also enable/enhance the Teaching/Learning Mission of the NE department at NCSU.

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Turbulent MHD flow modeling in annular linear induction pumps with validation experiments

US Dept. of Energy (DOE)(10/01/16 - 9/30/20)

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Turbulent MHD flow modeling in annular linear induction pumps with validation experiments

Sponsor: US Dept. of Energy (DOE)

Start Date: 10/01/16

End Date: 9/30/20

Abstract

This proposed scope of work seeks to advance EMP design efforts for liquid metal reactors by 1.) improving MHD modelling capability, particularly by incorporating turbulence modeling and investigating the MHD stability criteria, 2.) advancing multi-physics simulation of these systems leveraging the open source MOOSE platform that is already used in many nuclear reactor and reactor system efforts, and 3.) constructing a low barrier EMP test loop for model validation and instruction that can be easily replicated at low cost at other facilities.

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Microstructure Experiments-Enabled MARMOT Simulations of SiC/SiC-based Accident Tolerant Nuclear Fuel System

US Dept. of Energy (DOE)(10/01/16 - 9/30/19)

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Microstructure Experiments-Enabled MARMOT Simulations of SiC/SiC-based Accident Tolerant Nuclear Fuel System

Sponsor: US Dept. of Energy (DOE)

Start Date: 10/01/16

End Date: 9/30/19

Abstract

Following the nuclear accident at Fukushima, replacing the current Zr alloy-based cladding with silicon carbide composites has gained traction in the DOE as well as in the industry. The most recent generation of SiC/SiC composites (III), defined by near-stoichiometric chemical composition with reduced oxygen content, have been extensively tested for nuclear applications at ORNL and other places. Relatively new to the manufacturing landscape, the fabrication of SiC/SiC composites-based cladding system poses several challenges; recent evaluations have identified several important technological gaps such joining technology and steam oxidation. In the current work, an experimentally-validated simulation framework is proposed that is specifically addressing the key technology gap associated with steam oxidation attack and irradiation for the Triplex design with two types of composites – Tyranno SA3 and Hi-Nicalon Type S. The framework includes (i) Steam exposure tests-ORNL, (ii) Ion irradiation experiments-UTK/ORNL, (iii) 3D mapping of the microstructure using x-ray microscopy and mechanical tests-USC, and (iv) Microstructure evolution and chemical kinetics in MARMOT, phase field computer code developed at INL, under normal/accident conditions, and under different irradiated conditions, followed by validation against experimental results-NCSU/INL.

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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

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

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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Mechanistic and Validated Creep/Fatigue Predictions for Alloy 709 from Accelerated Experiments and Simulations

US Dept. of Energy (DOE)(10/01/15 - 9/30/19)

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Mechanistic and Validated Creep/Fatigue Predictions for Alloy 709 from Accelerated Experiments and Simulations

Sponsor: US Dept. of Energy (DOE)

Start Date: 10/01/15

End Date: 9/30/19

Abstract

As a promising candidate for fast reactor program, Alloy 709 possesses excellent high temperature thermo-mechanical properties. To support its qualification in the ASME code for Class 1 Components in Elevated Temperature Service (Section 3, Division 1, Subsection NH), we propose mechanistic methods for predicting creep and creep-fatigue deformation rates/strains based on accelerated in-situ and ex-situ tests, and mesoscale dislocation dynamics simulations. There are four research tasks in the proposed program. In Task 1, ex-situ creep and creep-fatigue experiments will be conducted by PD/PI (Murty, NCSU) at high temperatures and stresses followed by microstructure evaluations using high resolution transmission electron microscopy (HR-TEM), and Atom Probe Tomography (APT). To evaluate the time evolution of the microstructure, in Task 2, in-situ X-ray synchrotron (XRD) experiments will be conducted by PI-3 (Almer, ANL) and PI-2 (Kaoumi, USC) at the Advanced Photon Source (APS), along with in-situ TEM experiments by PI-2. Using the microstructure information from Tasks 1 and 2, mesoscale dislocation dynamics (DD) will be conducted by PI-4 (Eapen, NCSU) in Task 3 to follow the microstructure evolution at accelerated and reactor operating conditions. In Task 4, realistic continuum creep damage mechanics (CDM), informed by in-situ experiments and simulations on microstructure evolution, will be employed to predict the accumulated strains during creep and creep-fatigue at reactor operating conditions.

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Aging of Used Nuclear Fuel in Storage: Accelerated Characterization and Modeling of Performance

US Dept. of Energy (DOE)(12/15/11 - 12/15/15)

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Aging of Used Nuclear Fuel in Storage: Accelerated Characterization and Modeling of Performance

Sponsor: US Dept. of Energy (DOE)

Start Date: 12/15/11

End Date: 12/15/15

Abstract

This is a subcontract to TAMU (Dr. Sean McDevitt) with NCSU PIs to work on creep of spent fuel as a part of the proposal on aging of used nuclear fuel in storage being submitted to DOE/NEUP under solicitation: CALL FOR PROPOSALS (CFP) NO. NEUP-002-11 under Integrated Research Projects.

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