About

The work focuses on testing and development of computational tools to simulate near and long term Accident Tolerant Fuel (ATF) options, advanced fuels for LWR power uprate and advanced fuels for plutonium burning. Primarly testing facilities include prototypic LWR operating pressure, temperature and chemistry mechanical and corrosion capability, high temperature (<1500 C) air/steam oxidation column and low temperature (<300 C) burst facility. Primarily simulation tools usesd for the projects are MOOSE/BISON, FRAPCON and ABAQUS.

Non-MIT collaborators include Boise State University, Framatome, Free-Form-Fibers, Idaho National Laboratory, Los Alamos National Laboratory, Plasma Pros, Structural Integrity, Texas A&M, University of Florida, University of Wisconsin.

Near Term ATF Concepts: Coated Clad and Doped Fuel

Student: Yifeng Che (PhD), Mohammad Shahin (SM)

In order to estimate the economical impact under normal operation and additional grace period given by the ATF concpets data-driven simulations needs to be performed. The experimental work for validation include pressure tube and four-point bend crack propogation tests. The data will drive finite element simulations developed under INL's MOOSE/BISON framework.

Journal Publication: Wagih, M., Spencer, B., Hales, J., Shirvan, K., "Fuel performance of chromium-coated zirconium alloy and silicon carbide accident tolerant fuel claddings," Annals of Nuclear Energy, Vol 120, pp. 304-318, 2018.

Sevecek M., Gurgen A., Seshadri A., Che Y., Wagih M., Phillips B., Champagne V, Shirvan K., "Development of Cr Cold-Sprayed Fuel Cladding with Enhanced Accident Tolerance," Nuclear Engineering and Technology Journal, Vol 50, pp. 229-236, 2018.

Che, Y., Pastore, G., Hales, J., Shirvan, K., "Modeling of Cr2O3-doped UO2as a near-term accident tolerant fuel for LWRs using the BISON code," Nuclear Engineering and Design, Vol 337 pp. 271-278, 2018.

Long Term ATF Concepts: SiC/SiC Clad and Laser Printed Fuel-in-Fiber

Postdoc: Wei Li Student: Yanbin Deng (PhD), Arunkumar Seshadri (PhD), Mohammad Shahin (SM)

SiC/SiC fiber composite claddings provide high temperature oxidation and strength capability. The goal of our study is to investigate its application as a cladding or as a fuel matrix. For the cladding, multilayer CVD and fiber composite cladding along with different coatings (FeCrAl/Cr) is investigated. For fuel matrix, a laser CVD technique developed by Free-Form-Fibers is leveraged to deposit UN or U3Si2 fuel in cylinderical fiber shape, where every fuel fiber is coated with layers of carbon followed by SiC similar to TRISO fuel. The clear advantage of this appraoch is almsot 2 times the fuel packing fraction that combine with high density fuel allows for <6% enrich fuel loading for current LWRs. We utilize prototypic LWR test facitlties to measure the corrosion of SiC/SiC uncoated and coated claddings and fuel-in-fiber concept. Both surface characterization and strength testing of irradiated SiC/SiC samples from MITR are also performed. Our simulation work primarly uses FEA based fuel performance analysis with MOOSE/BISON and ABAQUS due to complex nature of the material architecture.

Journal Publication: Wagih, M., Spencer, B., Hales, J., Shirvan, K., "Fuel performance of chromium-coated zirconium alloy and silicon carbide accident tolerant fuel claddings," Annals of Nuclear Energy, Vol 120, pp. 304-318, 2018.

Deng, Y., Shirvan, K., Wu, Y., Su, G. "Probabilistic view of SiC/SiC composite cladding failure based on full core thermo-mechanical response," Journal of Nuclear Materials, Vol 507, pp. 24-37, 2018.

High Density Fuels: Nitride and Silicide

Post Doc: Wei Li Student: Mohammad Shahin (SM)

The current UO2 fuel has limited uranium density and thermal condutivity. Uranium nitride and silicide have a higher theoretical density and higher thermal conductivity than uranium oxide, allowing for lower fuel temperatures,longer cycle length and higher power densities with similar enrichment. The testing involves quantification of hydrothermal corrosion of these fuels. The analysis aims at finding the maximum thermal power and cycle length over a range of bundle geometries(rod diameter and pitch). Commerical reactor physics tools are utilized while MOOSE/BISON fuel performance suplemented by FRAPCON and ABAQUS is used for simulation.

Journal Publication: None

Fuels for LWR Power Uprate: Annular and Cruciform

Student: Yangbin Deng (Phd)

This work focuses on changing the fuel geometry while maintaining the operating conditions similar to current Pressurized Water Reactors (PWRs), in order to achieve significant power uprate. The Internally and eXternally cooled Annular Fuel (IXAF) is one such geometry that increases the heat transfer area of the fuel rod significantly and has shown to be able to increase the power output of existing PWRs by 50%. IXAF is cooled on both the external and internal surfaces, resulting in significantly lower average fuel temperatures, even at 50% higher power rating compared to equivalent solid pin design. The Helical Cruciform shaped Fuel (HCF) geometry uses the strategy of fins to increase the heat transfer area and also the twisted tapes approach to increase the swirl and intra-bundle mixing of the flow to increase margin to critical heat flux compared to the traditional cylindrical fuel rod bundle geometries.

Journal Publication: Shirvan K., "Numerical Investigation of the Boiling Crisis for Helical Cruciform-Shaped Rods at High Pressures," Journal of Multiphase Flow vol. 83, pp. 51-61, July, 2016.

Shirvan K., and M.S. Kazimi M., "Three Dimensional Considerations in Thermal-Hydraulics of Helical Cruciform Fuel Rods for LWR Power Uprates," Nuclear Engineering and Design, vol. 270, pp. 259-272, 2014.

Shirvan K., Hejzlar P., and Kazimi M.S., "The Design of A Compact Integral Medium Size PWR," Nuclear Engineering and Design vol. 243, pp. 393-403, Feb 2012.

Recent Finished Projects: MOX, Thoria, Additives

Student (Past): Yanin Sukjai (PhD)

Several phenomena at high temperature and high burnup regime such as fuel restructuring, plutonium redistribution, and fission product migration were modeled in FRAPCON for MOX and high burnup fuels. For Thoria fuels, updated fision gas reslease and thermal conductivity models for Thoria-Urania and Thoria-Plutonia fuels were developed in FRAPCON. Impact of BeO additive in fuel on fuel performance were also validated with FRAPCON.

Journal Publication: None