Abstract
Demonstration of hermetic SiC fiber–reinforced SiC matrix composite cladding under normal operating environments has been identified as one of the most critical feasibility issues for accident-tolerant fuel cladding in light-water reactors. This study provides critical experimental data needed for understanding the effects of irradiation on hermeticity. SiC composite and monolithic tubes were neutron-irradiated to 2 displacements per atom with and without a nominal radial heat flux of 0.6 MW/m2 to produce a simulated in-pile stress state for the normal operation of a light-water reactor. The through-thickness temperature gradient under irradiation results in a gradient in swelling, which causes a significant stress buildup. Such irradiation-induced stress was modeled using a commercial finite element analysis code. The radial heat flux–irradiation synergism was experimentally investigated by constructing a special irradiation capsule and evaluating the helium hermeticity of the specimens. The simulated stress state exhibited a near equi-biaxial tensile axial and hoop stress of ∼150 MPa at the inner surface of the SiC composite tube. This stress level is potentially beyond the matrix cracking stress. Degradation of hermeticity of the SiC composite tubes was observed after irradiation, indicating irradiation-induced cracking, whereas the irradiated monolithic SiC tubes remained hermetic. The results indicate that loss of hermeticity caused by radiation-induced microcracking is a potential issue for SiC composite cladding, depending on the magnitude of the temperature gradients. Coating the outer surface of the cladding was identified as a mitigation strategy that might overcome the cracking issue.
| Original language | English |
|---|---|
| Article number | 154784 |
| Journal | Journal of Nuclear Materials |
| Volume | 588 |
| DOIs | |
| State | Published - Jan 2024 |
Funding
This study was supported by the Advanced Fuels Campaign of the US Department Energy ( DOE ), Office of Nuclear Energy , and the Westinghouse Electric Corporation / General Atomics funding opportunity announcement program under contact DE-AC05–00OR22725 with ORNL , managed by UT Battelle, LLC . The irradiation experiments were also supported by the US DOE , Office of Nuclear Energy , under DOE Idaho Operations Office Contract DE-AC07–051D14517 as part of an award provided by the Nuclear Science User Facilities program . A portion of this research used resources at the HFIR , a DOE Office of Science User Facility operated by ORNL . Patricia Tedder, Stephanie Curlin, and Travis Dixon at ORNL contributed to the post-irradiation examinations. The authors wish to thank TS Byun, Tim Lach, and Erica Heinrich at ORNL and Sean Gonderman at General Atomics for reviewing and editing this manuscript. Weon-Ju Kim and Daejong Kim acknowledge support from the National Research Foundation of Korea ( NRF ) and a grant funded by the Korean government ( MSIT ) (No. 2017M2A8A4017642 ). This study was supported by the Advanced Fuels Campaign of the US Department Energy (DOE), Office of Nuclear Energy, and the Westinghouse Electric Corporation/General Atomics funding opportunity announcement program under contact DE-AC05–00OR22725 with ORNL, managed by UT Battelle, LLC. The irradiation experiments were also supported by the US DOE, Office of Nuclear Energy, under DOE Idaho Operations Office Contract DE-AC07–051D14517 as part of an award provided by the Nuclear Science User Facilities program. A portion of this research used resources at the HFIR, a DOE Office of Science User Facility operated by ORNL. Patricia Tedder, Stephanie Curlin, and Travis Dixon at ORNL contributed to the post-irradiation examinations. The authors wish to thank TS Byun, Tim Lach, and Erica Heinrich at ORNL and Sean Gonderman at General Atomics for reviewing and editing this manuscript. Weon-Ju Kim and Daejong Kim acknowledge support from the National Research Foundation of Korea (NRF) and a grant funded by the Korean government (MSIT) (No. 2017M2A8A4017642).
| Funders | Funder number |
|---|---|
| TS Byun, Tim Lach | |
| Travis Dixon at ORNL | |
| Westinghouse Electric Corporation | |
| Westinghouse Electric Corporation/General Atomics | |
| U.S. Department of Energy | DE-AC07–051D14517 |
| Office of Science | |
| Office of Nuclear Energy | |
| Oak Ridge National Laboratory | |
| UT-Battelle | |
| General Atomics | DE-AC05–00OR22725 |
| Ministry of Science, ICT and Future Planning | 2017M2A8A4017642 |
| National Research Foundation of Korea |
Keywords
- Accident tolerant fuel
- Composite
- Hermeticity
- Neutron irradiation
- SiC