Microstructural Analysis of an HT9 Fuel Assembly Duct Irradiated in FFTF to 155 Dpa at 443ºC

Journal Article · Tue Sep 01 00:00:00 EDT 2009 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2009.06.010· OSTI ID:968548

Sencer, Bulent H; Cole, James I; Kennedy, John R; Maloy, Stuart A; Garner, Frank A

The majority of published data on the irradiation response of ferritic/martensitic steels has been derived from simple free-standing specimens irradiated in experimental assemblies under well-defined and near-constant conditions, while components of long-lived fuel assemblies are more complex in shape and will experience progressive changes in environmental conditions. To insure that the resistance of HT9 to void swelling is maintained under more realistic operating conditions, this study addresses the radiation-induced microstructure of an HT9 ferritic/martensitic (F/M) steel hexagon duct that was examined following a six-year irradiation campaign of a fuel assembly in the Fast Flux Test Reactor Facility (FFTF). The calculated irradiation exposure and operating temperature of the duct location examined were ~155 dpa at ~443ºC. It was found that dislocation networks were contained predominantly a/2<111> Burgers vector. Surprisingly, for such a large irradiation dose, type a<100> interstitial loops were observed at relatively high density. Additionally, a high density of precipitation was observed. These two microstructural characteristics may have contributed to the rather low swelling level of 0.3%. It appears that the inherent swelling resistance of this alloy observed in specimens irradiated under non-varying experimental conditions is not significantly degraded compared to time-dependent variations in neutron flux-spectra, temperature and stress state that are characteristic of actual reactor components.

• Research Organization: Idaho National Lab. (INL), Idaho Falls, ID (United States)
• Sponsoring Organization: DOE - NE
• DOE Contract Number: DE-AC07-05ID14517
• OSTI ID: 968548
• Report Number(s): INL/JOU-09-15494; JNUMAM; TRN: US0904701
• Journal Information: Journal of Nuclear Materials, Vol. 393, Issue 2; ISSN 0022-3115
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
36 MATERIALS SCIENCE
ALLOYS
BURGERS VECTOR
DISLOCATIONS
DUCTS
FUEL ASSEMBLIES
INTERSTITIALS
IRRADIATION
MICROSTRUCTURE
NEUTRONS
PRECIPITATION
REACTOR COMPONENTS
SHAPE
STEELS
SWELLING
TEST REACTORS
Fast reactor
FFTF/MOTA
HT9
Microstrutural evolution
Void swelling