Ex-reactor testing for carbon loss from HT9 cladding in the presence of metal fuels
Year: 1987
Type: report
Abstract: U-10Zr fuel elements clad in HT9 and heated 1n an axial temperature gradient (500-700°C) were sectioned after 840 and 2325 h of heat treatment to look for carbon transfer from the HT9 cladding (0.2 wt % C) to the U-10Zr fuel (100 ppm C). The element had been encapsulated and sodium bonded in a nickel 200 capsule. It was assumed that carbon would not transfer between the Ni and HT9 tubes, at least from HT9 to N1200. Results obtained, however, showed that C did transfer from the Ni (0.08 wt %) to the HT9 cladding. There was no evidence of transfer from the HT9 to the fuel despite the extensive carburization of the HT9 cladding. There was therefore no strong tendency found for carbon to transfer from the HT9 cladding to the fuel to temperatures of 700°C and times of 2325 hours. Recent post-Irradiation testing of 3 at.% burnup X425 fuel elements support this finding. In addition, a literature review Indicates that carbon loss alone may not result in significant reductions in creep rupture life; recrystallization and grain growth at high temperatures resulting from loss of carbides causes the most significant strength reductions.
DOI: 10.2172/713759
Authors: G. D. Hudman, D. D. Keiser, D. L. Porter
Language: English
Subfield: Materials Chemistry
Field: Materials Science
Domain: Physical Sciences
Open Access status: green
OpenAlex ID: W2992245764