CORROSION OF ZIRCALOY IN CREVICES UNDER NUCLEATE BOILING CONDITIONS
The MSAR corrosion test loop was used to study the effect of VEE-shaped crevices on the corrosion resistance of Zircaloy under boiling heat-transfer conditions in both ammoniated and lithiated water at pH 10.5. Zircaloy-2 and Zircaloy-4 were compared under identical geometric, thermal, and hydraulic conditions. An effort was made to study the effect of iron oxide crud deposits on crevices of slightly variant geometry. The tests were run at 586 to 587 deg F (T/sub ast/ for a pressure of 1390 to 1400 psig). Water velocity was held in the range of 1 to 3 fps. Control was such that the onset of nucleate boiling should, on the average, have taken place at about the middle of the 5-in.long, thin- walled crevice specimens. Specimens were vertically oriented in an effort to mock up the most common orientation of flow channels in nuclear reactor cores. Heating of specimens was resistive with 60-cycle a-c, at a heat flux of 500,000 Btu/ft/sup 2/-hr The test in ammoniated water was operated successfully for 1000 hr. The test in lithiated water was subject to such drastic corrosion that it had to be terminated at a total exposure of 211 hr. Two factors tended to obscure the results of this test. First, the previously unsuspected release of alumina into loop water during the ammoniated-water test increased the ionic content of stagnant water in the crevices. Boehmite was deposited in these crevices during boiling. Second, the possible slow thermal dissociation of Teflon seals may have produced minute concentrations of fluorine-containing gases that cause accelerated attack on upstream bus bars. Under ammonia pH control, pitting of the Zircaloy specimens and white oxide formation were quite minor. After 1000 hr of testing, the maximum metal penetration found was two mils, observed at the closed edge of a Zircaloy-4 VEE. The maximum penetration depth observed in Zircaloy-2 was one mil. Under lithium hydroxide pH control, penetrations observed were 5.2 and 5.9 mils for Zircaloy-2 and Zircaloy-4, respectively, during the 211-hr test. Hydrogen pickup in the ammonia chemistry environment was negligible for both Zircaloy-2 and Zircaloy-4. Conversely, in the lithium hydroxide environment, drastic hydrogen embrittlement occurred. Extreme segregation of hydride platelets was demonstrated on all specimens exposed to lithiated water; 3100 ppm of hydrogen was found in the Zircaloy-4 VEE specimen compared with 1200 ppm for the Zircaloy-2 VEE in this boiling lithium- chemistry environment. Less corrosion, hydrogenation, pitting, and white oxide formation were observed in both tests on the precrudded, parallel-fuced crevice specimens than were seen on either of the comparable noncrudded VEE's. Whether this is attributable to the deposited synthetic crud or to the more open crevice design is not plain from the existing data. It was concluded that either Zircaloy-2 or Zircaloy-4 will undergo crevice corrosion in closed vertical channels under boiling heat transfer conditions at temperatures near 587 deg F. Much more drastic attack occurred in lithiated water than in ammoniated water. Under either condition, Zircaloy-4 appeared less resistant to attack than does Zircaloy-2. In the presence of lithium hydroxide, extreme hydrogen embrittiement took place with either Zircaloy. However, Zircaloy-4 was more strongly attacked. (auth)
- Research Organization:
- Knolls Atomic Power Lab., Schenectady, N.Y.
- DOE Contract Number:
- W-31-109-ENG-52
- NSA Number:
- NSA-16-019287
- OSTI ID:
- 4830227
- Report Number(s):
- KAPL-2203
- Resource Relation:
- Other Information: Orig. Receipt Date: 31-DEC-62
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
ACIDITY
ADSORPTION
ALUMINUM OXIDES
AMMONIA
BOILING
BRITTLENESS
CORROSION
CRACKS
DECOMPOSITION
DISTRIBUTION
EFFICIENCY
FILMS
FLUORINE
GASES
GRAIN BOUNDARIES
HEAT TRANSFER
HYDROGEN
HYDROGENATION
IRON OXIDES
LITHIUM HYDROXIDES
MATERIALS TESTING
PIPES
POLYMERS
PRESSURE
SEALS
STABILITY
TEMPERATURE
WATER
ZIRCALOY