Overview of the multifaceted activities towards development and deployment of nuclear-grade FeCrAl alloys

Conference · Fri Jul 01 04:00:00 EDT 2016

OSTI ID:22764064

Field, K. G.; Pint, B. A. ^[1]; Yamamoto, Y.; Terrani, K. A. ^[2]

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1. Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)
2. Fusion and Materials for Nuclear Systems Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

A large effort is underway under the leadership of US DOE Fuel Cycle Research program to develop advanced FeCrAl alloys as accident tolerant fuel (ATF) cladding to replace Zr-based alloys in light water reactors. The primary motivation is the excellent oxidation resistance of these alloys in high-temperature steam environments right up to their melting point (roughly three orders of magnitude slower oxidation kinetics than zirconium). A multifaceted effort is ongoing to rapidly advance FeCrAl alloys as a mature ATF concept. The activities span the broad spectrum of alloy development, environmental testing (high-temperature high-pressure water and elevated temperature steam), detailed mechanical characterization, material property database development, neutron irradiation, thin tube production, and multiple integral fuel test campaigns. Instead of off-the-shelf commercial alloys that might not prove optimal for the LWR fuel cladding application, a large amount of effort has been placed on the alloy development to identify the most optimum composition and microstructure for this application. The development program is targeting a cladding that offers performance comparable to or better than modern Zr-based alloys under normal operating and off-normal conditions. This paper provides a comprehensive overview of the systematic effort to advance nuclear-grade FeCrAl alloys as an ATF cladding in commercial LWRs. (authors)

• Research Organization: American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 22764064
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
36 MATERIALS SCIENCE
ACCIDENT-TOLERANT NUCLEAR FUELS
ALUMINIUM ALLOYS
CHROMIUM ALLOYS
CLADDING
COMPARATIVE EVALUATIONS
FUEL CYCLE
IRON ALLOYS
MELTING POINTS
MICROSTRUCTURE
NEUTRONS
OXIDATION
RESEARCH PROGRAMS
STEAM
TEMPERATURE RANGE 0400-1000 K
TESTING
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCONIUM
ZIRCONIUM BASE ALLOYS