About this Abstract |
Meeting |
2023 TMS Annual Meeting & Exhibition
|
Symposium
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Mechanical Behavior of Nuclear Reactor Materials and Components III
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Presentation Title |
Mechanical Behavior of Bare and Cr Coated Zirconium Claddings During Burst Testing via In-situ Strain Measurements |
Author(s) |
Samuel Bell, Mackenzie Ridley, Kenneth Kane, Ben Garrison, Tim Graening, Nathan Capps |
On-Site Speaker (Planned) |
Samuel Bell |
Abstract Scope |
Cr coated zirconium alloys are a leading concept to replace the incumbent bare zirconium fuel claddings in light-water reactors. Cr coatings have demonstrated greater high temperature steam oxidation resistance, as well the potential for improved mechanical response during accident scenarios. Before this concept can be widely deployed, a better understanding of coated cladding behavior during accident scenarios is necessary. Recent efforts to improve burst testing, an established method of assessing cladding in simulated accident conditions, have integrated digital image correlation techniques to measure in-situ strain during rapid heating of pressurized cladding materials until rupture. In this work, digital image correlation techniques were applied to monitor both Cr coated and bare Zircaloy-4 cladding segments during burst testing. In-situ strain measurements and mechanical behavior of bare and Cr coated Zircaloy-4 will be compared. |
Proceedings Inclusion? |
Planned: |
Keywords |
Nuclear Materials, Mechanical Properties, Other |