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Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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Spent fuel transfer project completed at INL
Work crews at Idaho National Laboratory have transferred 40 spent nuclear fuel canisters into long-term storage vaults, the Department of Energy’s Office of Environmental Management has reported.
G. P. Airey, A. R. Vaia, R. G. Aspden
Nuclear Technology | Volume 55 | Number 2 | November 1981 | Pages 436-448
Technical Paper | Materials | doi.org/10.13182/NT55-436
Articles are hosted by Taylor and Francis Online.
Inconel 690 is an austenitic nickel base alloy that has been considered for use as steam generator tubing in pressurized water reactors. It has a composition comparable to the currently used Inconel 600 apart from a higher chromium content (30%). Inconel 690 was evaluated for stress corrosion cracking (SCC) resistance in deaerated sodium hydroxide solutions and all-volatile treatment (AVT) environments, and the results are compared with those from thermally treated Inconel 600. The effect of such metallurgical variables as grain size and thermal treatment on SCC resistance was investigated. The caustic SCC resistance of Inconel 690 was significantly improved, compared to the mill-annealed product, when thermally treated over a wide range of temperatures [649 to 871°C (1200 to 1600°F)] and times (1 to 30 h). The SCC resistance of thermally treated Inconel 690 was found to be excellent when exposed to deaerated sodium hydroxide at 343°C (650°F) and 316°C (600°F). There was an increase in caustic SCC susceptibility with decreasing grain size in mill-annealed tubing. However, the corresponding thermally treated tubing showed no caustic SCC susceptibility. No evidence of SCC degradation has been found in either mill-annealed or thermally treated Inconel 690 exposed to high temperature AVT and “pure water” environments.