Cannon, N. Scott
Senior engineer, Westinghouse Hanford Company, Richland, WA
Senior scientist, Westinghouse Hanford Company, Richland, WA
Hamilton, Margaret L.
Research scientist, Battelle Memorial Institute, Richland, WA
Pages: 10 Published: Jan 1990
Simulated transient tests were performed on sections of HT9 fast-reactor fuel pin cladding irradiated to a fast fluence of nearly
A slight reduction of strength was observed in irradiated cladding, particularly at 110°C/s, when compared with transient results from unirradiated HT9 control specimens; however, this strength reduction did not correlate with either fluence or irradiation temperature. A small reduction of ductility was also observed for irradiated cladding failing at temperatures above 800°C at the lower heating rates (0.56 or 5.6°C/s); irradiated cladding was generally more ductile at 110°C/s than unirradiated HT9 cladding.
The HT9 cladding results were compared with similar transient data obtained previously from 20% cold-worked type 316 stainless steel (316 SS) cladding. In the unirradiated state, this austenitic cladding is stronger and less ductile than HT9 cladding. However, the 316 SS cladding undergoes a significant loss of strength and ductility during irradiation when in contact with oxide fuel through a mechanism labeled the fuel adjacency effect (FAE). The FAE is believed to be the result of liquid metal embrittlement caused by fission products. The HT9 fuel-pin cladding remained as strong or stronger than the 316 SS cladding when irradiated in contact with fuel, and showed no evidence of the FAE up to the high fluences reported here. The ductility of the irradiated HT9 fuel-pin cladding remained significantly greater than that of irradiated 316 SS cladding.
radiation, irradiation, nuclear fuel cladding, transient, mechanical properties, strength, ductility, HT9
Paper ID: STP49487S