Assessing thermo-mechanical performance of ThO₂ and SiC clad light water reactor fuel rods with a modular simulation tool

This work has focused on evaluating the thermo-mechanical performance of two innovative nuclear fuel designs: the ThO2-based resource-renewable boiling water reactor (RBWR) and silicon carbide (SiC) cladding in pressurized water reactors (PWRs). Since these designs employ non-traditional fuel rod materials and/or operating conditions, the fuel behavior and limits lie outside of current experience and must be appropriately assessed. The RedTail fuel performance code was developed to support early-stage innovative fuel designs using established models within a modular code structure. Additionally, the applied assumptions have been selected to allow for fast run times. This versatility and speed allow the RedTail fuel performance code to be applied towards broad scoping studies of new fuel rod concepts. With its use of (ThUPu)O2 and high cladding fluence, new physical models were applied to describe the ThO2-based RBWR fuel rods. The effects of lattice strain induced phonon scattering and semi-conductive electron transport on the thermal conductivity for this ternary fuel mixture were modeled. A new model has also been created to reconstruct the radial distribution of the fission rate in the fuel. Lastly, the effects of high fast neutron fluence on the accelerated oxidation and hydrogen pickup by Zircaloy-2 have been investigated and incorporated. In developing the ThO2-based RBWR concept, three designs have been proposed based on different goals and applied assumptions. With the exception of hydrogen uptake, none of the predicted parameters exceeded their limitations. However, the application of hydrogen-based cladding accident limits was found to be restricting. A sensitivity study revealed that in order to retain non-zero accident limits, an advanced cladding is required. Because of its lower corrosion rate and high strength in high temperature steam, composite ceramic SiC cladding has the potential to improve the response to severe accidents. Applicable SiC material properties have been investigated, and new swelling and thermal conductivity models have been applied to treat radiation damage effects. In addition, Weibull failure mechanics are applied to CVD SiC. The fuel performance of every SiC clad pin in a PWR core has been evaluated to find the core-wide leaker risk. To account for uncertainty in potential SiC cladding designs 24 separate cases were evaluated along with a comparison to ZIRLO®. The three-layer SiC architecture is found not to be a viable option due to large shutdown stresses. The two-layer design, however, shows much lower failure risk. This study indicates that significant opportunities exist to optimize core design to improve SiC cladding performance.

Department dc:contributor.department

Massachusetts Institute of Technology. Department of Nuclear Science and Engineering.

Grantor dc:publisher

Massachusetts Institute of Technology

Year dc:date.issued

2015

Author dc:creator

• Mieloszyk, Alexander James

Advisor dc:contributor.advisor

• Neil E. Todreas.

• Nuclear Science and Engineering.

Statement dc:rights

• M.I.T. theses are protected by copyright. They may be viewed from this source for any purpose, but reproduction or distribution in any format is prohibited without written permission. See provided URL for inquiries about permission.

Licence dc:rights.uri

• http://dspace.mit.edu/handle/1721.1/7582

Language dc:language.iso

eng

Handle dc:identifier.uri

http://hdl.handle.net/1721.1/103660

OAI identifier oai:identifier

oai:dspace.mit.edu:1721.1/103660