Preliminary Investigation of Zircaloy-4 as a Research Reactor Cladding Material

Castle, Brian K

Idaho National Laboratory

关键词: Zircaloy;    Cladding;    Aluminum;    22 General Studies Of Nuclear Reactors Aluminum;   

DOI  :  10.2172/1056014
RP-ID  :  INL/EXT-12-25814
RP-ID  :  DE-AC07-05ID14517
RP-ID  :  1056014

美国|英语

来源: UNT Digital Library

【 摘 要 】

As part of a scoping study for the ATR fuel conversion project, an initial comparison of the material properties of Zircaloy-4 and Aluminum-6061 (T6 and O-temper) is performed to provide a preliminary evaluation of Zircaloy-4 for possible inclusion as a candidate cladding material for ATR fuel elements. The current fuel design for the ATR uses Aluminum 6061 (T6 and O temper) as a cladding and structural material in the fuel element and to date, no fuel failures have been reported. Based on this successful and longstanding operating history, Zircaloy-4 properties will be evaluated against the material properties for aluminum-6061. The preliminary investigation will focus on a comparison of density, oxidation rates, water chemistry requirements, mechanical properties, thermal properties, and neutronic properties.