The stress corrosion cracking of reactor pressure vessel steel in high temperature water

J. Congleton, T. Shoji, R. N. Parkins

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104 Citations (Scopus)


Slow strain rate stress corrosion tests have been performed on a nuclear reactor pressure vessel steel in pure water at temperatures from 150 to 288°C and at various oxygen contents. The variation of cracking susceptibility correlates with the electrochemical potential, and cracking occurs for this temperature range if the rest potential is greater than about -0.2 V(SHE). It was confirmed that the oxygen in the water acts as a chemical potentiostat, which drives the steel into or out of the cracking range, by showing that cracking could be prevented in 45 ppm O2 water at 288°C by applying cathodic polarization. Cracking could also be induced in < 5 ppb O2 content water by applying anodic potentials. Large IR drops occurred in the potentiostatically controlled tests and it is probable that the potential need only be moved slightly from the cracking potential to induce a significant reduction in the susceptibility to cracking. The cracking potential correlates well with the Fe3O4Fe2O3 boundary for high temperature potential-pH data. It is probable that the mechanism of stress corrosion cracking requires that both Fe2O3 and Fe3O4 can form in the crack enclave. By replotting the potential-pH data in the form of the required oxygen for a given potential against temperature, it is shown that the oxygen content for inducing cracking agrees closely with the Fe3O4-Fe2O3 boundary at temperatures from 150°C to 288°C and that in this temperature range cracking is more likely to be associated with an anodic dissolution mechanism than by hydrogen-assisted cracking.

Original languageEnglish
Pages (from-to)633-650
Number of pages18
JournalCorrosion Science
Issue number8-9
Publication statusPublished - 1985

ASJC Scopus subject areas

  • Chemistry(all)
  • Chemical Engineering(all)
  • Materials Science(all)


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