Corrosion of Zircaloy-2 by pH 10 LiOH in heated crevices
Both the inner and outer tubes of the N-Reactor fuel elements will have self supports spot welded to the lateral heat-transfer surface of the element. A crevice a few mils thick will exist around the weld between the support tab and the cladding. Because of the heat flux through the cladding at this point and the insulating effect of the support tab, the temperature in this crevice will be higher than that on the free surface away from the support. This can result in boiling in the crevice leading to concentration of LiOH (or impurities in the water) to a level where it can cause severe corrosion of the Zircaloy-2 cladding. The tests described in this report were conducted to determine whether such attack might be encountered in N-Reactor.
- Research Organization:
- General Electric Co., Richland, WA (United States). Hanford Atomic Products Operation
- Sponsoring Organization:
- USDOE, Washington, DC (United States)
- DOE Contract Number:
- AC06-76RL01830
- OSTI ID:
- 10145450
- Report Number(s):
- HW--77611; ON: DE94010508
- Country of Publication:
- United States
- Language:
- English
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