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

Abstract

As part of a scoping study for the Advanced Test Reactor (ATR) fuel conversion project, an initial comparison of the material properties of Zircaloy-4 (Zr-4) and aluminum-6061 (T6 and O-temper) is performed to provide a preliminary evaluation of Zr-4 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 and this design is proven to have a high degree of reliability. Because of the successful and longstanding operating history of aluminum-6061 (Al-6061) cladding in the ATR, Zr-4 properties will be evaluated against Al-6061 material properties. The preliminary investigation will focus on a comparison of densities, oxidation rates, water chemistry requirements, mechanical properties, thermal properties, and neutronic properties. Revision 0 9. Conclusions/Recommendations:In the event that aluminum becomes a nonviable option for the ATR fuel cladding, a more detailed analysis of the resulting thermal and structural changes, associated with a change in fuel cladding, should be performed. This assertion is predominantly based on the following: •The thermal conductivity of Zr-4 is significantly lower than Al-6061, which may necessitate operating the fuel at a higher centerline temperature and may result in non-beneficial changes to the operating parameters of the reactor.• Replacing the Al-6061 cladding with Zr-4 would increase the weight of the core by approximately 265.6 kg in addition to any increase from the fuel.