SYMPOSIA PAPER Published: 09 February 2018
STP87019850004

Fracture Resistance of Two Ferritic Stainless Steels After Intermediate Temperature Irradiation

Source

Two plates of Alloy HT-9 and one plate of modified Alloy 9Cr-lMo were irradiated at temperatures ranging from 93 to 300°C to study alloy resistance to radiation-induced embrittlement for fusion reactor applications. The neutron fluences were ~1 × 1020 neutrons (n)/cm2 (E > 1.0 MeV); assessments included the effects of alloy composition (12Cr versus 9Cr) and melt processing (Electroslag Refining (ESR) versus Argon Oxygen Decarburization (AOD)) on radiation resistance. Fracture resistance change was judged from pre versus postirradiation notch ductility and dynamic fracture toughness properties determined, respectively, with standard Charpy-V (Cv) and fatigue precracked Charpy-V (PCCv) tests.

A significant difference in radiation resistance was observed for the two alloys. The dissimilarity can be attributed to their respective compositions or microstructures or both. Comparable radiation resistance was found for ESR versus AOD plates from the same (parent) melt. Irradiation at 93°C elevated the brittle/ductile transition of the Alloy HT-9 to the simulated service temperature. In contrast, an appreciable toughness reserve against additional fluence was exhibited by both alloys after irradiation at 149°C (or higher).

The data indicate a general agreement of Cv and PCCv methods in their independent assessment of irradiation embrittlement. However, the alloys showed a very marked difference in the relative temperature ranges of the Cv and PCCv brittle/ductile transitions wherein modified 9Cr-lMo but not HT-9 appeared to have a marked sensitivity to test specimen notch acuity. Scanning electron microscope (SEM) examinations of the modified 9Cr-lMo specimens showed predominantly cleavage failure in PCCv tests at temperatures where Cv specimens failed predominantly in shear.

New data for Alloy HT-9 weld heat affected zone material and AISI Type 308-16 weld deposit materials irradiated at 288°C are also presented.

Author Information

Hawthorne, J., Russell
Naval Research Laboratory, Washington, DC, US
Reed, James, R.
Naval Research Laboratory, Washington, DC, US
Sprague, James, A.
Naval Research Laboratory, Washington, DC, US
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Developed by Committee: E10
Pages: 580–604
DOI: 10.1520/STP87019850004
ISBN-EB: 978-0-8031-7679-9
ISBN-13: 978-0-8031-0592-8