CORROSION BEHAVIOR OF FUEL ELEMENTS WITH U-2 w/o Zr CORE CLAD WITH ZIRCALOY- 2. Status Report
Further corrosion tests have been performed on sections from power tubes with a U--2 wt.% Zr core and 15-mil Zircaloy-2 cladding. Emphasis is on tests with a gross defect inserted in the outer cladding. These tests include cooling to simulate reactor shut-down at fixed intervals after attainment of a set level of corrosion at 660 F (349 C). Results presented for such tests in water and steam show that the consequences of an in-pile failure are not likely to be disastrous either in terms of uranium release or physical damage to the tube. As- extruded material is attacked less rapidly by 660 F (349 C) water than heat treated; the pattern of corrosion and damage is also affected by heat treatment. Quantitative results for heat-treated material are sensitive to the cooling rate snd to the location of the defect. Results are also presented for other corrosion tests, notably tests of tube sections with their transversely cut faces exposed. Such tests were applied to various questions such as the effect of hydrogen concentration during the corrosion test or the effect of modified heat treatments on the core alloy. (auth)
- Research Organization:
- Nuclear Metals, Inc., Concord, Mass.
- NSA Number:
- NSA-13-021192
- OSTI ID:
- 4234175
- Report Number(s):
- NMI-4388
- Country of Publication:
- United States
- Language:
- English
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