Haggag, Fahmy M.
Development staff member, Oak Ridge National Laboratory, Oak Ridge, TN
Corwin, William R.
Program manager, Oak Ridge National Laboratory, Oak Ridge, TN
Alexander, David J.
Metallurgist, Oak Ridge National Laboratory, Oak Ridge, TN
Nanstad, Randy K.
Group leader, Oak Ridge National Laboratory, Oak Ridge, TN
Pages: 12 Published: Jan 1990
The potential for stainless steel cladding to improve the fracture behavior of an operating nuclear reactor pressure vessel, particularly during certain overcooling transients, may depend greatly on the properties of the irradiated cladding. Therefore, three-wire stainless steel cladding irradiated at temperatures and to fluences relevant to power reactor operation was examined. Postirradiation testing results show that, in the test temperature range from − 125 to 288°C, the yield strength increased by 8 to 30%, and ductility insignificantly increased, while there was almost no change in the ultimate tensile strength. All cladding exhibited ductile-to-brittle transition behavior during Charpy impact testing, owing to the dominance of delta-ferrite failures at low temperatures. On the upper shelf, the energy was reduced (owing to irradiation exposure) 15 and 20%, while the lateral expansion was reduced 43 and 41% at 2 and
radiation damage, three-wire series-arc welds, stainless steels, pressure vessels, neutron irradiation, degradation, tension tests, toughness, Charpy impact
Paper ID: STP49460S