Tensile and Corrosion Behavior of Zr705 for Nuclear Hydrogen Generation

Document Type

Article

Publication Date

7-15-2006

Publication Title

Materials Science and Engineering A

Volume

427

Issue

1-2

First page number:

320

Last page number:

326

Abstract

Zr705 has been evaluated for its metallurgical and corrosion properties for application in nuclear hydrogen generation using a thermochemical process that involves the decomposition of hydroiodic acid at temperatures up to 400 °C. The results indicate that the tensile strength was gradually reduced with increasing temperature. However, the failure strain was enhanced up to a critical temperature (200 °C) followed by its reduction beyond it, possibly due to the dynamic strain aging effect. As to the cracking susceptibility in an acidic solution, no failure was observed under constant-load and self-loaded conditions. However, enhanced ductility was observed in slow-strain-rate (SSR) testing in an identical environment at elevated temperatures. The application of external potential during SSR testing enhanced the cracking susceptibility. The critical potentials obtained in the electrochemical polarization study became more active with increasing temperature. Ductile failures, characterized by dimples, were noted in all failed specimens.

Keywords

Applied potential; Fractography; Stress corrosion; Stress corrosion cracking; Zirconium alloys – Fatigue; Zirconium alloys – Fracture; Zr705

Disciplines

Materials Science and Engineering | Mechanical Engineering | Mechanics of Materials | Metallurgy | Nuclear Engineering

Language

English

Permissions

Use Find in Your Library, contact the author, or interlibrary loan to garner a copy of the item. Publisher policy does not allow archiving the final published version. If a post-print (author's peer-reviewed manuscript) is allowed and available, or publisher policy changes, the item will be deposited.

UNLV article access

Search your library

Share

COinS