Silicon Carbide Performance as Cladding for Advanced Uranium and Thorium Fuels for Light Water Reactors

Report Date: August 2016
Appendices: No

Executive Summary

In Phase I of the Lockheed Martin and Center for Advanced Nuclear Energy System (CANES) collaboration, the Thoria-Plutonia mixed-oxide (MOX) fuel exhibited superior performance compared to Urania-Plutonia MOX fuel. Phase I also concluded that burning plutonia in MOX fuels is more efficient as the amount of energy extracted for unit volume of fuel is increased (e.g. increase in burnup). To that end, this work focuses on technologies that can enable achieving high burnup fuel.

The Thoria fuel thermomechanical models developed in Phase I study were developed based on data from the light water breeder program in the 1970s and modeled using FRAPCON-MIT fuel performance tool. In this work, the validation database is expanded and more thorough comparison of the fuel thermal conductivity and fission gas release is made with experimental database that includes recent data from CANDU reactors. After the models are updated and validated, the effect of using thoria-plutonia fuel in lieu of urania-plutonia fuel for plutonium burning is investigated for Zircaloy and silicon carbide cladding. Such analysis is performed by simulating a standard LWR fuel rod during typical LWR conditions.

In order to achieve high burnup, reactivity control, fuel and cladding performance needs to be quantified and meet fuel licensing standards. Specifically, cladding mechanical performance and water-side corrosion along with fuel internal pressure are typically the main variables limiting the burnup of a nuclear fuel rod. In order to improve cladding mechanical performance, carbon nano-tubes (CNT) were mixed with Zriconium powder and subjected to ion irradiation to simulate reactor conditions. To address water-corrosion concerns, SiC cladding or coating can be utilized, however, there are mechanical implications by going from a ductile to a more brittle cladding material. Lastly, in order to address fuel internal pressure concerns of high burnup fuel, beryllium-oxide (BeO) can be mixed with the fuel to increase its thermal conductivity and thus decrease fission gas release that leads to lower internal pressure.

Program:     ANP : Advanced Nuclear Power Program

Type:     TR 

RPT. No.: 167