Keeping Fast Reactor Steel in Shape
Advanced oxide dispersion strengthened (ODS) ferritic alloys are potential top candidate materials for components that need exceptional resistance to radiation-induced swelling and creep.
In fast-neutron reactors, fuel is sealed in ~7 millimeter diameter steel tubes called cladding. When a high-energy "fast" neutron strikes an atom in the steel, it can knock the atom out of place, like a cue ball striking another billiard ball. This leaves two types of damage in the metal: an empty spot where the atom was, and the displaced atom wedged between other atoms. Over time, these defects typically drive undesirable rearrangement of the microstructure, potentially reducing the life of the cladding.
Oxide-dispersion strengthened (ODS) ferritic steels—iron-based alloys with fine, nanometer-size oxide particles dispersed in the metal—were identified in the late 1970s as promising fast-reactor fuel cladding materials due to their excellent resistance to thermal creep that degrades fuel cladding when pushed to high temperatures to achieve better conversion of heat to electricity. Current advanced fast reactor concepts also call for extended fuel burn to extract more energy, allowing for more economical electricity production and single core nuclear reactors that never need to be refueled over a set lifetime. This requires the fuel cladding to withstand up to 500+ displacements per atom (dpa) of exposure.
PNNL was the first national laboratory to pursue the development of ODS ferritic alloys for fast reactor applications in the 1980s. Researchers selected a ferritic alloy base due to its inherently high resistance to radiation-induced swelling. Once such alloy, called MA957, showed excellent resistance to creep and swelling after neutron irradiation to 100 dpa.
In order to assist in the creation of an optimized ODS ferritic alloy for the US-DOE Fuel Cycle R&D Program, scientists at PNNL set out to explore the high dose radiation resistance of two ODS ferritic alloys—MA957 and a newer alloy called 14YWT. Because it is impractical to achieve 500+ dpa from the available nuclear test reactors as part of an alloy development program, heavy-ion irradiation was used as a surrogate. This technique allows for high doses to be rapidly and economically obtained but does cause some small differences in microstructural evolution in ferritic steels when compared to neutron irradiation, so irradiations were performed on materials with known swelling response to a neutron irradiation environment for comparison. Whereas two non-ODS ferritic/martensitic steels irradiated in the same heavy-ion irradiation facility showed 10 percent or more swelling by 500 dpa exposure at 450°C, MA957 exhibited only 4.5 percent swelling and 14YWT exhibited only 1.5 percent swelling, showing this this class of alloy can provide much greater resistance to swelling, which is typically the life limiting degradation mode for fast reactor fuel cladding.
Microstructural observations of the irradiated ODS alloys by transmission electron microscopy (TEM) suggest that the oxide particles and very small grain size achieved in these materials play a major role in controlling the amount of irradiation-induced swelling. Additional studies using TEM and atom probe tomography are under way to characterize the microstructural stability of these materials with these results to contribute to further development of this class of alloy and their likely selection as the fuel cladding material for next generation high dose fast reactors.
This research was funded by the Fuel Cycle Research and Development program, sponsored by the U.S. DOE Office of Nuclear Energy. Collaboration with the Kharkov Institute of Physics and Technology in the Ukraine was made possible by the CRDF organization.
Published: June 20, 2016