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ASTM Selected Technical Papers
Zirconium in the Nuclear Industry: 20th International Symposium
Editor
Suresh K. Yagnik
Suresh K. Yagnik
Symposium Chairperson and STP Editor
1
Electric Power Research Institute (EPRI)
,
Palo Alto, CA,
US
Search for other works by this author on:
Michael Preuss
Michael Preuss
Symposium Chair and STP Editor
2
The University of Manchester Manchester
,
GB
;
Monash University
,
Clayton/Melbourne,
AU
Search for other works by this author on:
ISBN:
978-0-8031-7737-6
No. of Pages:
928
Publisher:
ASTM International
Publication date:
2023

Fuel cladding tubes made of zirconium alloys, are subjected in reactor to a complex loading history under nominal operating conditions. Furthermore, they exhibit a complex deformation behavior resulting from irradiation-induced growth, irradiation creep and thermal creep. For design and safety requirements, empirical models are usually used. To have robust physically based mechanical simulations, a self-consistent polycrystalline model has been developed. This model takes into account the various phenomena occurring at the grain scale, such as irradiation-induced growth and irradiation creep. Moreover, this model takes into account the crystallographic texture of the material and the mechanical interactions between grains, depending on their orientation. Furthermore, this model is able to handle complex mechanical loading. This model is first shown to reproduce well an experimental database of in-reactor deformation of zirconium alloys. Thanks to the polycrystalline nature of this model, the effect of grain shape and creep mechanisms at the grain scale on the simulated data have been studied in detail. Next, this polycrystalline model has been introduced into a 1D finite element method code, allowing the computation of stress and strain gradients through a thin cladding tube during a complex mechanical loading. This approach opens the way to physically based mechanical calculations at the component scale.

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