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

Deducing Phonon Scattering from Normal Mode Excitations

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

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

Ion hopping and constrained Li diffusion pathways in the superionic state of antifluorite Li2O

Annamareddy, A., & Eapen, J. (2017), Entropy, 19(5).

Low dimensional string-like relaxation underpins superionic conduction in fluorites and related structures

Annamareddy, A., & Eapen, J. (2017), Scientific Reports, 7. https://doi.org/10.1038/srep44149

View all publications via NC State Libraries

Grants

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 - 9/30/19)

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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: 9/30/19

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 - 9/30/19)

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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: 9/30/19

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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One-Dimensional Nanostructures for Neutron Detection

US Dept. of Energy (DOE)(9/28/11 - 9/30/14)

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One-Dimensional Nanostructures for Neutron Detection

Sponsor: US Dept. of Energy (DOE)

Start Date: 9/28/11

End Date: 9/30/14

Abstract

Compact size, low operating voltage, fast charge-collection speed and high sensitivity are important measures of neutron detectors. Other measures include low radiation damage and compatibility with different substrates. Recent advances in materials processing show that low dimensional structures such as nanotubes and nanowires have superior electrical sensitivity relative to bulk materials. This proposal aims to develop high efficiency, pixel array neutron detectors using boron carbon nitride nanotubes.

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Consortium for Advanced Simulations for Light Water Reactors (CASL) - Oak Ridge National laboratory

US Dept. of Energy (DOE)(11/30/-1 - 9/30/19)

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Consortium for Advanced Simulations for Light Water Reactors (CASL) - Oak Ridge National laboratory

Sponsor: US Dept. of Energy (DOE)

Start Date: 11/30/-1

End Date: 9/30/19

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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