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Published March 1965 | Version v1
Journal article

Corrosion studies of reactor materials by high temperature and high pressure water loop, (3). Corrosion of Zr alloys (2)

  • 1. Japan Atomic Energy Research Institute, Reactor Mater. Res. Br., Tokai, Ibaraki (Japan)

Description

Subsequent to the previous lower temperature experiment the corrosion of Zr-2 and -4 coupons in flowing water, at 2.5 and 12.0 m/s, was studied under 300° and 320℃. The weight gain of coupons by dynamic corrosion was represented as the following function of time, similarly to the case of static corrosion: logΔW = log K+n log t where however the constants K and n are both dependent on the flow conditions and temperature of water (K generally increases with temperature at same flow rate). From a comparison of the constants under different flow conditions, it is suggested that the corrosion behavior of the coupons is more similar to static corrosion when exposed to flow at 2.5 m/s than at 12.0 m/s, in which case the dissolution of the oxide film could be observed by measuring the impedance of the oxide film. The effects on corrosion of initially dissolved oxygen concentration and of replacement of the samples were found to be negligible. Hydrogen pick-up by the coupons was little influenced by the flow conditions, and the difference of pick-up between Zr-2 and Zr-4 was much reduced at 320℃. A large amount of crud deposition was experienced on the coupons heated to 320℃, though coupons replaced during an experiment under the same temperature were found almost quite free of crud. (author)

Additional details

Additional titles

Original title (Japanese)
高温高圧水ループによる原子炉材料の腐食の研究,(3).Zr合金の腐食,(2)

Identifiers

Publishing Information

Journal Title
Nippon Genshiryoku Gakkai-Shi
Journal Volume
7
Journal Issue
3
Series
雑誌名:日本原子力学会誌
Journal Page Range
p. 148-157
ISSN
0004-7120

Optional Information

Notes
10 refs., 11 figs., 2 tabs., 2 photos