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

Comparison of glancing-angle scatterings on different materials in a high aspect ratio plasma etching process using molecular dynamics simulation

Du, Y., Kruger, F., Nam, S. K., Lee, H., Yoo, S., Eapen, J., … Shannon, S. (2022), JOURNAL OF VACUUM SCIENCE & TECHNOLOGY A, 40(5). https://doi.org/10.1116/6.0002008

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

Decoding ionic conductivity and reordering in cation-disordered pyrochlores

Annamareddy, A., & Eapen, J. (2021), PHILOSOPHICAL TRANSACTIONS OF THE ROYAL SOCIETY A-MATHEMATICAL PHYSICAL AND ENGINEERING SCIENCES, 379(2211). https://doi.org/10.1098/rsta.2019.0452

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

View all publications via NC State Libraries

Grants

Simulating Yttrium Hydride for Microreactor Moderators

Idaho National Laboratory (was Idaho National Engineering Lab)(5/03/23 - 8/14/23)

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Simulating Yttrium Hydride for Microreactor Moderators

Sponsor: Idaho National Laboratory (was Idaho National Engineering Lab)

Start Date: 5/03/23

End Date: 8/14/23

Abstract

The principal aim of the proposed project is to simulate a promising metal hydride (YH2) using computer simulations. The PI will work with Dr. Yu, who is an expert in computational materials science and Dr. Cinbiz, who is an expert on experimental techniques and analysis. Density functional theory (DFT) and ab initio simulations will be conducted using the electronic structure simulation package VASP. Statistical mechanical analysis will then be performed to understand the transport of hydrogen in YH2 at different temperatures. The objective of the project is to extract the atomistic mechanisms of hydrogen transport in YH¬2 and help in interpreting the data obtained at INL using experimental techniques.

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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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Quantifying the Dynamic and Static Porosity/Microstructure Characteristics of Irradiated Graphite through Multi-technique Experiments and Mesoscale Modeling

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

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Quantifying the Dynamic and Static Porosity/Microstructure Characteristics of Irradiated Graphite through Multi-technique Experiments and Mesoscale Modeling

Sponsor: US Dept. of Energy (DOE)

Start Date: 10/01/21

End Date: 9/30/24

Abstract

We propose a joint experimental-computational approach to probing and quantifying the porosity and microstructure characteristics of irradiated nuclear graphite grades and their influence on dimensional changes and turnaround behavior as well as mechanical properties. Our chief focus is on quantifying both the static and dynamic porosity/crack characteristics in various graphitic phases (filler particles, binders, quinoline insoluble particles) in medium, fine and superfine grained irradiated graphite grades (NBG-17/18, PCEA, IG-110/430, G347A, POCO (ZXF-5Q/AXF-5Q) and HOPG) through several experimental techniques. HOPG is included as a highly ordered/oriented reference grade while the POCO grades, which do not have a binder phase, serve as an intermediate form between HOPG and other nuclear grades. The samples are neutron irradiated at temperatures ranging from 300°C to 1500°C, and the doses span from 1 to 21 dpa, approximately. Additionally, our international collaborating team at the University of Manchester will investigate graphite grades that are oxidized; we thus will cover a spectrum of irradiation, thermal and oxidative conditions in this project.

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Nuclear Energy University Programs - Fellowship and Scholarship Support North Carolina State University

US Dept. of Energy (DOE)(8/01/21 - 8/31/33)

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Nuclear Energy University Programs - Fellowship and Scholarship Support North Carolina State University

Sponsor: US Dept. of Energy (DOE)

Start Date: 8/01/21

End Date: 8/31/33

Abstract

This proposal is a request to be eligible to participate in DOE's Fellowship and Scholarship Support program where NE eligible students can apply for scholarships and fellowships per the Funding Opportunity Announcement. This is up to a 10 year cooperative agreement program and may run up to 13 years to accredited US Colleges and Universities for the US Department of Energy's Office of Nuclear Energy Fellowship and Scholarship Awards.  

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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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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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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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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 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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