Assessment of Corrosion Resistance of Candidate Alloys for Accident Tolerant Fuel Cladding under Reactor Conditions
The environment of a light water reactor (LWR) core is a combination of high temperatures, high pressures, high neutron and gamma fluxes, mechanical stresses and chemical aggressive coolants. All of these factors combined can induce changes in the microstructure of the fuel and cladding that are very difficult to predict in a systematic fashion [1] and are major challenge for the safety and life extension of current reactors. Throughout reactor transients and accidents, the cladding may experience deterioration caused by a temperature increase, oxidation embrittlement [1], [2], or mechanical interaction with the fuel caused by stress [3]. These events may lead to cracking or rupture of the cladding, causing the release of fission products into the coolant. Such events have been observed at the Three Mile Island and Fukushima accidents. To avoid such occurrences in LWRs, the accident tolerant fuels (ATF) program was initiated to focus on the replacement of zirconium-based alloys with materials that exhibit slower steam oxidation kinetics. This project focuses on several iron-based alloys such as T91, APMT, MA956, experimental Fe-Cr alloys and one experimental nanofeatured alloy (NFA). Experiments have been conducted in both PWR primary water at 320°C and BWR normal water chemistry at 288°C, spanning a large range in electrochemical corrosion potential (ECP). Samples were exposed to either proton irradiation (University of Michigan) or electron irradiation (Notre Dame Radiation Laboratory) to independently assess the roles of displacement damage or radiolysis on the corrosion rate, oxide thickness, morphology, structure and resistivity. Post-irradiation characterization of various regions of the electronirradiated samples was completed by means of microscopy (Notre Dame Integrated Imaging Facility) and spectroscopy (Notre Dame Materials Characterization Facility) techniques to provide high resolution information regarding the oxide layer present on the surface of the material.
- Research Organization:
- Notre Dame University
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE), Fuel Cycle Technologies (NE-5)
- DOE Contract Number:
- NE0008272
- OSTI ID:
- 1474325
- Report Number(s):
- 3
- Journal Information:
- Microscopy and Microanalysis, Vol. 23 (Suppl 1), Issue S1; Conference: Microscopy and Microanalysis, August 6-10, 2017, St. Louis, Missouri.; ISSN 1431-9276
- Publisher:
- Microscopy Society of America (MSA)
- Country of Publication:
- United States
- Language:
- English
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