Evaluation on the Effect of Composition on Radiation Hardening and Embrittlement in Model FeCrAl Alloys

Technical Report · Fri Sep 18 00:00:00 EDT 2015

DOI:https://doi.org/10.2172/1253237· OSTI ID:1253237

Field, Kevin G. ^[1]; Briggs, Samuel A. ^[1]; Edmondson, Philip ^[1]; Hu, Xunxiang ^[1]; Littrell, Kenneth C. ^[1]; Howard, Richard ^[1]; Parish, Chad M. ^[1]; Yamamoto, Yukinori ^[1]

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1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

This report details the findings of post-radiation mechanical testing and microstructural characterization performed on a series of model and commercial FeCrAl alloys to assist with the development of a cladding technology with enhanced accident tolerance. The samples investigated include model alloys with simple ferritic grain structure and two commercial alloys with minor solute additions. These samples were irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) up to nominal doses of 7.0 dpa near or at Light Water Reactor (LWR) relevant temperatures (300-400 C). Characterization included a suite of techniques including small angle neutron scattering (SANS), atom probe tomography (APT), and transmission based electron microscopy techniques. Mechanical testing included tensile tests at room temperature on sub-sized tensile specimens. The goal of this work was to conduct detailed characterization and mechanical testing to begin establishing empirical and/or theoretical structure-property relationships for radiation-induced hardening and embrittlement in the FeCrAl alloy class. Development of such relationships will provide insight on the performance of FeCrAl alloys in an irradiation environment and will enable further development of the alloy class for applications within a LWR environment. A particular focus was made on establishing trends, including composition and radiation dose. The report highlights in detail the pertinent findings based on this work. This report shows that radiation hardening in the alloys is primarily composition dependent due to the phase separation in the high-Cr FeCrAl alloys. Other radiation induced/enhanced microstructural features were less dependent on composition and when observed at low number densities, were not a significant contributor to the observed mechanical responses. Pre-existing microstructure in the alloys was found to be important, with grain boundaries and pre-existing dislocation networks acting as defect sinks, resulting in variations in the observed microstructures after irradiation. Dose trends were also observed, with increasing radiation dose promoting changes in the size and number density of the Cr-rich α' precipitates. Based on the microstructural analysis, performed tensile testing, and prior knowledge from FeCr literature it was hypothesized that the formation of the Cr-rich α' precipitates could lead to significant radiation-induced embrittlement in the alloys, and this could be composition dependent, a result which would mirror the trends observed for radiation-induced hardening. Due to the limited database on embrittlement in the FeCrAl alloy class after irradiation, a series of radiation experiments have been implemented. The overarching point of view within this report is the radiation tolerance of FeCrAl is complex, with many mechanisms and factors to be considered at once. Further development of the FeCrAl alloy class for enhanced accident tolerant applications requires detailed, single (or at least limited) variable experiments to fully comprehend and predict the performance of this alloy in LWRs. This report has been submitted as fulfillment of milestone M2FT-15OR0202321 titled, Summary report on the effect of composition on the irradiation embrittlement of Gen 1 ATF FeCrAl for the Department of Energy Office of Nuclear Energy, Advanced Fuel Campaign of the Fuel Cycle R&D program.

• Research Organization: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE)
• DOE Contract Number: AC05-00OR22725
• OSTI ID: 1253237
• Report Number(s): ORNL/TM-2015/518; AF5810000; NEAF278; M2FT- 15OR0202321; TRN: US1601368
• Country of Publication: United States
• Language: English

Related Subjects

36 MATERIALS SCIENCE
TEMPERATURE RANGE 0400-1000 K
RADIATION DOSES
RADIATION HARDENING
GRAIN BOUNDARIES
ACCIDENT-TOLERANT NUCLEAR FUELS
EMBRITTLEMENT
DISLOCATIONS
IRRADIATION
PRECIPITATION
TEMPERATURE RANGE 0273-0400 K
FUEL CANS
EVALUATION
PERFORMANCE
DEFECTS
SINKS
TESTING
VARIATIONS
CHEMICAL COMPOSITION
IRON BASE ALLOYS
CHROMIUM ALLOYS
ALUMINIUM ALLOYS
TERNARY ALLOY SYSTEMS
MECHANICAL TESTS