Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

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• 作者：Robert Montgomery^a ; ^{robert.montgomery@pnnl.gov" class="auth_mail" title="E-mail the corresponding author} ; Carlos Tomé ; ^b ; ^{tome@lanl.gov" class="auth_mail" title="E-mail the corresponding author} ; Wenfeng Liu^c ; ^{wenfeng.liu@anatech.com" class="auth_mail" title="E-mail the corresponding author} ; Alankar Alankar^d ; ^{alankar.alankar@iitb.ac.in" class="auth_mail" title="E-mail the corresponding author} ; Gopinath Subramanian^e ; ^{gopinath.subramanian@usm.edu" class="auth_mail" title="E-mail the corresponding author} ; Christopher Stanek^b ; ^{stanek@lanl.gov" class="auth_mail" title="E-mail the corresponding author}
• 关键词：Nuclear fuel modeling ; Microstructure deformation modeling ; Zirconium alloy ; Irradiation growth ; Irradiation creep ; Plasticity
• 刊名：Journal of Computational Physics
• 出版年：2017
• 出版时间：1 January 2017
• 年：2017
• 卷：328
• 期：Complete
• 页码：278-300
• 全文大小：2656 K

文摘

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.