Effect of controlled potential on SCC of nuclear waste package container materials
The slow-strain-rate (SSR) test technique was used to evaluate the susceptibility of Titanium (Ti) Gr-7 (UNS R52400) and Ti Gr-12 (UNS R53400) to stress corrosion cracking (SCC). Ti Gr-7 and Ti Gr-12 are two candidate container materials for the multi-barrier package for nuclear waste. The tests were done in a deaerated 90 C acidic brine (pH {approx} 2.7) containing 5 weight percent (wt%) sodium chloride (NaCl) using a strain rate of 3.3 x 10{sup -6} sec{sup -1}. Before being tested in the acidic brine, specimens of each alloy were pulled inside the test chamber in the dry condition at ambient temperature. Then while in the test solution, specimens were strained under different cathodic (negative) controlled electrochemical potentials. These controlled potentials were selected based on the corrosion potential measured in the test solution before the specimens were strained. Results indicate that the times to failure (TTF) for Ti Gr-12 were much shorter than those for Ti Gr-7. Furthermore, as the applied potential became more cathodic, Ti Gr-12 showed reduced ductility in terms of percent reduction in area (%RA) and true fracture stress ({sigma}{sub f}). In addition, TTF and percent elongation (%El) reached the minimum values when Ti Gr-12 was tested under an impressed potential of -1162 mV. However, for Ti Gr-7, all these ductility parameters were not significantly influenced by the changes in applied potential. In general, the results of hydrogen analysis by secondary ion mass spectrometry (SIMS) showed increased hydrogen concentration at more cathodic controlled potentials. Optical microscopy and scanning electron microscopy (SEM) were used to evaluate the morphology of cracking both at the primary fracture face and the secondary cracks along the gage section of the broken tensile specimen. Transgranular secondary cracks were observed in both alloys possibly resulting from the formation of brittle titanium hydrides due to cathodic charging. The primary fracture face was characterized by dimpled microstructure indicating ductile failure.
- Research Organization:
- Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
- Sponsoring Organization:
- USDOE Office of Civilian Radioactive Waste Management (RW) (US)
- DOE Contract Number:
- W-7405-ENG-48
- OSTI ID:
- 13784
- Report Number(s):
- UCRL-JC-135596; TRN: US0110752
- Resource Relation:
- Conference: Corrosion/2000 Conference, Orlando, FL (US), 03/26/2000--03/31/2000; Other Information: PBD: 1 Aug 1999
- Country of Publication:
- United States
- Language:
- English
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