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Replacement Zircaloy for Silicon Carbide as Fuel Cladding Material in PWR – A Neutronic Evaluation

Published online by Cambridge University Press:  13 February 2015

Rochkhudson B. de Faria ,
Felipe Torres ,
Fabiana B. A. Monteiro  and
Claubia Pereira
Rochkhudson B. de Faria*
Affiliation:
Departamento de Engenharia Nuclear – Escola de Engenharia Universidade Federal de Minas Gerais Avenida Antônio Carlos, 6627, Pampulha. 31270-901 – Belo Horizonte, Tel/Fax: 55-31-34096662, MG, Brazil.
Felipe Torres
Affiliation:
Departamento de Engenharia Nuclear – Escola de Engenharia Universidade Federal de Minas Gerais Avenida Antônio Carlos, 6627, Pampulha. 31270-901 – Belo Horizonte, Tel/Fax: 55-31-34096662, MG, Brazil.
Fabiana B. A. Monteiro
Affiliation:
Departamento de Engenharia Nuclear – Escola de Engenharia Universidade Federal de Minas Gerais Avenida Antônio Carlos, 6627, Pampulha. 31270-901 – Belo Horizonte, Tel/Fax: 55-31-34096662, MG, Brazil. Instituto Nacional de Ciências e Tecnologia de Reatores Nucleares Inovadores/CNPq, Brazil
Claubia Pereira
Affiliation:
Departamento de Engenharia Nuclear – Escola de Engenharia Universidade Federal de Minas Gerais Avenida Antônio Carlos, 6627, Pampulha. 31270-901 – Belo Horizonte, Tel/Fax: 55-31-34096662, MG, Brazil. Instituto Nacional de Ciências e Tecnologia de Reatores Nucleares Inovadores/CNPq, Brazil
*
*rochkhudson@ufmg.br
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Abstract

Silicon carbide (SiC) has a potential to replacement zircaloy as fuel cladding material due to its high temperature tolerance, chemical stability and low neutron affinity. These characteristics may improve the economic and safety of nuclear reactors. Previous work has examined the possible use of SiC as a fuel cladding material in a PWR (Pressurized Water Reactor) environment. However, the advantage thermo mechanical and neutronic analysis replacement zircaloy cladding is not clear. Literature reviews has been done to predict the thermo mechanical behavior of SiC in high temperatures. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. This codes system is widely accepted and used worldwide for safety analysis and criticality of nuclear reactors has been utilized to model a typical fuel element of a PWR. It was used the CSAS6 and TRITON modules. The goals are to evaluate the behavior of the infinite multiplication factor (kinf) and neutron flux using SiC as a fuel cladding material.

Keywords

Nuclear materialsalloycoating
Type
Articles
Information
MRS Online Proceedings Library (OPL) , Volume 1769: Symposium 6E – Materials for Nuclear Applications , 2015 , IMRC 2014 6E-15
Copyright
Copyright © Materials Research Society 2015 

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References

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